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Sample records for reactor metodo para

  1. 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)

  2. A digital method for period measurements in a nuclear reactor; Um metodo digital para medidas de periodo em um reator nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Mundim, Sergio Gorretta

    1971-02-15

    The present paper begins by giving a theoretical treatment for the nuclear reactor period. The conventional method of measuring the period is analysed and some previously developed digital methods are described. The paper criticises the latter, pointing out some deficiencies which the proposed process is able to eliminate. All errors connected with this process are also analysed. The paper presents suitable solutions to reduce them to a minimum. The total error is found to he less than the error presented by the other methods described. A digital period meter is designed with memory resources and an automatic scaler changer. Integrated circuits specifications are used in it. Real time experiments with nuclear reactors were made in order to check te validity of the method. The data acquired were applied to a simulated digital period meter implemented in a general purpose computer. The nuclear part of the work was developed at the 'Comissao Nacional de Energia Nuclear' and the simulation work was dane at the 'Departamento de Calculo Cientifico' of COPPE, which also advised the author in the completion of this thesis. (author)

  3. A digital method for period measurements in a nuclear reactor; Um metodo digital para medidas de periodo em um reator nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Mundim, Sergio Gorretta

    1971-02-15

    The present paper begins by giving a theoretical treatment for the nuclear reactor period. The conventional method of measuring the period is analysed and some previously developed digital methods are described. The paper criticises the latter, pointing out some deficiencies which the proposed process is able to eliminate. All errors connected with this process are also analysed. The paper presents suitable solutions to reduce them to a minimum. The total error is found to he less than the error presented by the other methods described. A digital period meter is designed with memory resources and an automatic scaler changer. Integrated circuits specifications are used in it. Real time experiments with nuclear reactors were made in order to check te validity of the method. The data acquired were applied to a simulated digital period meter implemented in a general purpose computer. The nuclear part of the work was developed at the 'Comissao Nacional de Energia Nuclear' and the simulation work was dane at the 'Departamento de Calculo Cientifico' of COPPE, which also advised the author in the completion of this thesis. (author)

  4. Industrial Ultrasonic Inspection of Stainless-Steel Claddings for the EL4 Reactor; Controle Industriel par Ultrasons des Gaines en Acier Inoxydable du Reacteur EL4; Promyshlennyj kontrol' obolochechnykh trub iz nerzhaveyushchej stali reaktora dlya EL4 s pomoshch'yu ul'trazvukovogo metoda; Metodos Ultrasonicos para Control Industrial de las Vainas de Acero Inoxidable del Reactor EL4

    Energy Technology Data Exchange (ETDEWEB)

    Prot, A. C.; Foulquoer, H. E.; Peyrot, J. P. [Centre d' Etudes Nucleaires de Saclay (France)

    1965-09-15

    del metodo a utilizar representa un proceso delicado, cuyas consideraciones fundamentales se exponen en el presente trabajo. Una vez elegido el metodo y puesto a punto en el laboratorio, surgen dos nuevos problemas: Transposicion a escala industrial. Necesidad de tener siempre presente la calidad que puede alcanzarse en la industria, en relacion con normas de aceptacion definidas de manera mas o menos arbitraria. En la practica, ello obliga a realizar un estudio estadistico sobre partidas de tubos de diversos origenes y clasificarlos teniendo en cuenta umbrales de aceptacion de distintos grados de severidad. Como se ve en el trabajo, el numero de tubos a controlar es muy superior al previsto inicialmente. Este hecho indujo a estudiar una maquina de control automatico, capaz de satisfacer al mismo tiempo las exigencias de la cantidad y las propias del tipo de control seleccionado; estas ultimas son por lo general de orden mecanico y requieren una construccion especialmente esmerada. El conjunto de estas consideraciones llevo a concebir una maquina capaz de satisfacer sin dificultad las necesidades de una cadena de fabricacion'de elementos combustibles. Las posibilidades de esta maquina estan estrechamente ligadas a las caracteristicas del material descontrol escogido, sobre todo al rendimiento de los circuitos electronicos correspondientes a los aparatos de control por metodos ultrasonicos y al de los transductores utilizados. Se deduce del presente estudio, por otra parte, que el material corriente no responde al problema sino de manera muy imperfecta, y que se debe encarar desde ya el proyecto de un aparato especial para este tipo de control. (author) [Russian] Uluchshenie rabochih harakteristik reaktorov trebuet primenenija tshhatel'no razrabotannyh i strogo kontroliruemyh materialov. Odnim iz aspektov jetogo kontrolja javljaetsja kachestvo ispol'zuemyh pokrytij dlja trub, mehanicheskoe sostojanie kotoryh predstavljaet soboj sushhestvennyj faktor rentabel

  5. New Methods and Facilities for the Measurement of Physical Properties of Reactor Components and Irradiated Materials; Nouveaux Procedes et Instruments de Mesure des Proprietes Physiques des Elements de Reacteur et des Matieres Irradiees; Novye metody i sredstva izmereniya fizicheskikh s vojstv komponentov reaktora i obluchennykh materialov; Nuevos Metodos y Equipos para Medir Propiedades Fisicas de Componentes de Reactor y de Materiales Irradiados

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-09-15

    radiactivas a temperaturas comprendidas entre 20 y 1 000 Degree-Sign C. Los autores discuten el control de calidad de metales no ferrosos por medida de la conductividad electrica aplicando corrientes de Foucault. Se describe un instrumento para medir la conductividad electrica de metales no ferrosos sin utilizar elementos en contacto. Se explica la correlacion entre la conductividad electrica y las caracteristicas de esfuerzo-deformacion de metales y aleaciones no ferrosos. Se presta especial atencion a la determinacion de estas propiedades en muestras pequenas. Se describe un dispositivo para medicion directa a distancia en la zona radiactiva del reactor. Se examina la relacion existente entre conductividad electrica y la dosis de radiacion. Se describe un instrumento para medir la permeabilidad, la remanencia y la fuerza coercitiva en funcion de la-solicitacion mecanica y la deformacion elastica y plastica, como tambien de la dosis de radiacion. Se expone un metodo para medir la variacion de las propiedades magneticas en funcion de la solicitacion elastica y la deformacion plastica. Se discuten los efectos de la irradiacion sobre la permeabilidad y la fuerza coercitiva. Se describe un instrumento para medicion rapida e indicacion directa de la permeabilidad de componentes de acero inoxidable. Se explica la correlacion entre permeabilidad y contenido de {Delta} - ferrita. Se exponen las mediciones del porcentaje de {Delta} - ferrita en soldaduras practicadas en tubos de acero inoxidable, y de las precipitaciones de {Delta}-ferrita en funcion de la deformacion plastica (forjado de los elementos combustibles). (author) [Russian] Opisan pribor dlja avtomaticheskogo izmerenija i registracii modulja Junga, modulja sdviga i kojefficienta zatuhanija kak funkcii temperatury i vremeni. Izmerenie modulja Junga proizvoditsja s pomoshh'ju vozbuzhdennyh obrazcov razlichnyh razmerov pri ih estestvennoj chastote. Izmerenie kojefficienta zatuhanija proizvoditsja po svobodnomu zatuhaniju

  6. Atlas de aves: Un metodo para documentar distribucion y seguir poblaciones

    Science.gov (United States)

    Robbins, C.S.; Dowell, B.A.; Dawson, D.K.; Alvarez-Lopez, Humberto; Kattan, Gustavo; Murcia, Carolina

    1988-01-01

    Los Atlas de Aves son proyectos nacionales o regionalies para trazar en mapas la distribucion en reproduccion de cada especie de ave. Ese procedimiento se esta usando en Europa, Australia, Nueva Zelanda, Norteamerica, y partes de Africa. El tama?o de los cuadrados varia de medio grado de latitud y Iongitud hasta 5 x 5 km. El trabajo de campo de cada proyecto exige aproxlmadamente cinco a?os, pero los aficionados pueden llevar a cabo la mayor parte del trabajo. Es posible almacenar los resultados en un computador personal. Hay muchos beneficios: (I) se presenta la distribucion corriente de las aves de la nacion, del estado, o de la Iocalidad; (2) se desarrolla nueva informacion especialmente sobre especies raras o en peligro; (3) se descubren areas que tienen una avlfauna sobresaliente o habitats raros y ayuda a su proteccion, (4) se documentan cambios de dlstribucion; (5) se pueden usar para documentar cambios de poblacion, especialmente en los tropicos donde otros metodos son mas dificiles de usar porque hay muchas especies y no hay muchos observadores calificados en la identificacion de sonidos de las aves; (6) son proyectos buenos de investigacion para estudiantes graduados; (7) los turistas y los jefes de excursiones de historia natural pueden contribuir con muchas informaciones

  7. A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods; Um estudo de arquiteturas de hardware para aplicacao em sistemas digitais de protecao de reatores nucleares - metodos de analise de confiabilidade e seguranca

    Energy Technology Data Exchange (ETDEWEB)

    Benko, Pedro Luiz

    1997-07-01

    A study of digital hardware architectures, including experience in many countries, topologies and solutions to interface circuits for protection systems of nuclear reactors is presented. Methods for developing digital systems architectures based on fault tolerant and safety requirements is proposed. Directives for assessing such conditions are suggested. Techniques and the most common tools employed in reliability, safety evaluation and modeling of hardware architectures is also presented. Markov chain modeling is used to evaluate the reliability of redundant architectures. In order to estimate software quality, several mechanisms to be used in design, specification, and validation and verification (V and V) procedures are suggested. A digital protection system architecture has been analyzed as a case study. (author)

  8. Methods in nuclear reactors calculations; Metodos de calculo en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G

    1966-07-01

    Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P{sub l}; B{sub l}; M{sub l}; S{sub n} and discrete ordinates approximations. (Author)

  9. Method to allow the estimation of heat transfer coefficients in solar stills; Metodo para determinar coeficientes locales de transferencia de calor en destiladores solares

    Energy Technology Data Exchange (ETDEWEB)

    Rubio Cerda, Eduardo; Porta Gandara, Miguel A [CIBNOR, Mexico D.F (Mexico); Fernandez Zayas, Jose Luis [UNAM Mexico, D.F. (Mexico)

    2000-07-01

    This work reports an experimental method that allows to estimate the heat transfer coefficients in the neighborhood of walls or flat plates subject to convective transport phenomena. This method can be applied to a great variety of thermal systems since it is based on the knowledge of the border condition for the temperature at the surface of the plate, and the temperature profile that characterize the dimensionless coefficient of heat transfer in the fluid, according to its definition given by the Nusselt number. The approach of this work are the foundations of the method and the system that has been developed to apply it, that incorporates automatic acquisition equipment for continuos monitoring of the information and elements to control the parameters of interest. In addition, the experimental cavities on which the method will be evaluated are discussed, considering two different scales, as well as experiments in cavities filled with air, and with a mixture of air and steam water, as is the case for solar distillation. [Spanish] En este trabajo se presenta un metodo que permite determinar de manera experimental coeficientes de transferencia de calor por conveccion. Este metodo puede ser aplicado a una gran variedad de sistemas termicos ya que se fundamenta en el conocimiento de la condicion de frontera para la temperatura en la superficie de la placa, y del perfil de temperaturas que caracteriza el coeficiente adimensional de transferencia de calor en el fluido, de acuerdo a la definicion de este, dada por el numero de Nusselt. El trabajo que aqui se reporta esta enfocado a la fundamentacion del metodo y al equipamiento que se ha desarrollado para instrumentarlo, que incorpora equipos automaticos de adquisicion continua de informacion y elementos de control para los parametros de interes. Se presentan ademas, las cavidades experimentales sobre las que sera evaluado el metodo, que considera dos escalas diferentes, asi como experimentos en cavidades llenas de aire

  10. Microscale adaptation of the potentiometric method with ion-selective electrode for the quantification of fluoride; Adaptacion a microescala del metodo potenciometrico con electrodo ion selectivo para la cuantificacion de fluoruro

    Energy Technology Data Exchange (ETDEWEB)

    Guevara Ruiz, Paulina; Ortiz Perez, Maria Deogracias [Laboratorio de Bioquimica, Facultad de de Medicina, Universidad Autonoma de San Luis Potosi, San Luis Potosi, San Luis Potosi, (Mexico)]. E-mail: mdortiz@uaslp.mx

    2009-05-15

    Similarly to other countries, ground water from Mexico is naturally polluted by fluoride. The main effects of fluoride at typical ground water concentrations are dental fluorosis, neurological deficits and reproductive disorders. In order to verify that the fluoride concentration is within the allowed guideline in Mexico (NOM 127 and 201), it is important to monitor fluoride levels in water and commercial beverages. The aim of this work is to develop a modification of the standard potentiometric method for fluoride determination in water, in order to reduce costs and amount of potentially toxic waste substances. Both methods were validated, the standard potentiometric method with the ion selective electrode and the microscale modification proposed in this paper. The methods were compared using statistic tests and graphics, followed by the comparison of 125 samples of commercial bottled water sold in the city of San Luis Potosi. Optimal results were obtained for the validation of both methods, and the microscale modification showed statistically identical results to those obtained with the standard method in all samples of bottled water. The microscale modification is a good alternative for fluoride assessment in water and beverages, and it represents a 95 % reduction of costs and chemical waste. [Spanish] En varios paises, incluido Mexico se presenta una contaminacion natural con fluoruro en agua subterranea; los principales efectos en la salud observados en poblacion expuesta a concentraciones mayores al valor permisible (que en Mexico es de 1.5 mg/L) son la fluorosis dental y esqueletica, asi como dano reproductivo y neurologico. En varios estados de la republica Mexicana, este problema es aun desconocido, de ahi la necesidad de evaluar las concentraciones de fluoruro en agua de consumo en varias comunidades. Asi, el objetivo de este trabajo es desarrollar un metodo a microescala para la determinacion de fluoruro en agua, que al reducir la cantidad de reactivo y

  11. Method of identifying the friction of rotors using the wavelet transform; Metodo para identificar el rozamiento de rotores utilizado la transformada wavelet

    Energy Technology Data Exchange (ETDEWEB)

    Jauregui Correa, Juan Carlos; Rubio Cerda, Eduardo; Gonzalez Brambila, Oscar [CIATEQ, A.C., Queretaro (Mexico)

    2007-11-15

    The modern processes of signal analysis that measure mechanical vibrations are based on the fast transform of Fourier (FFT), nevertheless, this method is not able to identify transient phenomena nor of nonlinear nature. Although many efforts have been made to try to identify these phenomena in the frequency spectra, it is not possible to correlate the spectra with the physical characteristics of this type of phenomena. Within these phenomena on the rubbing of a rotor against the housing or trunnion of a bearing, this phenomenon has a nonlinear behavior, as it is demonstrated in this paper. In the first part a method based on the of signal analysis type wavelets is presented and how this technique can be used to predict transient and nonlinear phenomena. Once defined the method, its application in the identification of the friction of rotors is demonstrated. With this, one demonstrates that the method presented in this paper allows to also identifying in real time the rubbing phenomenon and also that it can be used as an of analysis technique in the preventive maintenance systems. [Spanish] Los procesos modernos de analisis de senales que miden vibraciones mecanicas se basan en la transformada rapida de Fourier (FFT por sus siglas en ingles), sin embargo, este metodo no es capaz de identificar fenomenos transitorios ni de naturaleza no lineal. A pesar de que se han hecho muchos esfuerzos para tratar de identificar estos fenomenos en los espectros de frecuencia, no es posible correlacionar el espectro con las caracteristicas fisicas de este tipo de fenomenos. Dentro de estos fenomenos sobre el rozamiento de un rotor contra la carcasa o munon de una chumacera, este fenomeno tiene un comportamiento no lineal, como se demuestra en este trabajo. En la primera parte se presenta un metodo basado en el analisis de senales tipo wavelets y como esta tecnica puede utilizarse para predecir fenomenos transitorios y no lineales. Una vez definido el metodo, se demuestra su

  12. METODO OPTICO PARA MEDIR EL COMPORTAMIENTO DE UNA LINEA DE TRANSMISION.

    OpenAIRE

    ASAHI KODAMA, TAKESHI EDUARDO; ASAHI KODAMA, TAKESHI EDUARDO

    1998-01-01

    El objetivo de esta tesis es el de establecer un método de manera de poder medir en forma óptica la corriente y el voltaje que circulan en una línea de transmisión. Para ello, se utilizan los fenómenos de Faraday y Kerr en líneas de transmisión que tienen 57p.

  13. METODO PROPUESTO PARA LA PREDICCION DE TENSIONES ADMISIBLES EN ZAPATAS CIMENTALDAS EN ARENAS

    OpenAIRE

    DIAZ SEGURA; EDGAR GIOVANNY; DIAZ SEGURA; EDGAR GIOVANNY

    2010-01-01

    La evidencia experimental ha demostrado que la mayoría de las metodologías actuales no son capaces de predecir las cargas de diseño de zapatas en suelos no cohesivos con un grado aceptable de precisión. En el presente estudio se propone un método sencillo y realista para estimar la carga de diseño de zapatas rígidas cimentadas en arenas sometidas a una carga estática vertical. Se adoptó como criterio de diseño el propuesto por Terzaghi et al. (1996), basado en limitar a 16 mm e...

  14. METODO PROPUESTO PARA LA PREDICCION DE TENSIONES ADMISIBLES EN ZAPATAS CIMENTADAS EN ARENAS

    OpenAIRE

    DIAZ SEGURA, EDGAR GIOVANNY; DIAZ SEGURA, EDGAR GIOVANNY

    2010-01-01

    La evidencia experimental ha demostrado que la mayoría de las metodologías actuales no son capaces de predecir las cargas de diseño de zapatas en suelos no cohesivos con un grado aceptable de precisión. En el presente estudio se propone un método sencillo y realista para estimar la carga de diseño de zapatas rígidas cimentadas en arenas sometidas a una carga estática vertical. Se adoptó como criterio de diseño el propuesto por Terzaghi et al. (1996), basado en limitar a 16 mm el asentamiento ...

  15. Um metodo para modelagem de exceções em desenvolvimento baseado em componentes

    OpenAIRE

    Patrick Henrique da Silva Brito

    2005-01-01

    Resumo: Devido a grande popularização do Desenvolvimento Baseado em Componentes (DBC), ele vem sendo empregado inclusive no desenvolvimento de sistemas computacionais críticos. O emprego do DBC na construção de sistemas confiáveis evidencia a necessidade de se desenvolver componentes de software que sejam robustos e que possuam uma garantia maior do seu funcionamento correto. Tratamento de exceções é uma técnica bastante conhecida para a verificação e tratamento de erros em sistemas de softwa...

  16. Metodo para montaje permanente de huevos de helmintos enteroparasitos Definitive preservation of helminthic eggs

    Directory of Open Access Journals (Sweden)

    Victor Muñoz

    1990-04-01

    Full Text Available Se comunican resultados obtenidos empleando Medio de Hoyer para el montaje de huevos de helmintos enteroparásitos, destinado a preparaciones para colecciones docentes y/o de investigación. La utilización de esta técnica en muestras fecales conteniendo huevos de A. lumbricoides, T. trichiura, Uncinaria sp., Taenia sp., Diphyllobothrium sp., H. nana, H. diminuta y F. hepática, permitió la correcta observación de ellos en lecturas iniciadas a las 24 horas y mantenidas hasta 180 días después.The results using the Hoyer method for examining eggs of helminths enteroparasites are presented. This method is particularly swited for teaching and on research purposes. Using this technique in fecal sample containing eggs of A.lumbricoides, T. trichiura, Uncinaria sp., Taenia sp., Diphyllobothrium sp., H. nana, H. diminuta and F. hepática allowed the correct identification of then after 24 hours up to 180 days after the samples were obtained.

  17. A reverse method to estimate initial temperatures in geothermal reservoirs; Un metodo inverso para estimacion de la temperatura inicial de yacimientos geotermicos

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Gutierrez, Alfonso [Instituto de Investigaciones Electricas, Gerencia de Geotermia, Cuernavaca, Morelos (Mexico)]. E-mail: aggarcia@iie.org.mx; Ramos Alcantara, Jose R. [Centro Nacional de Investigacion y Desarrollo Tecnologico, Departamento de Ingenieria Mecanica, Cuernavaca, Morelos (Mexico); Arellano Gomez, Victor M. [Instituto de Investigaciones Electricas, Gerencia de Geotermia, Cuernavaca, Morelos (Mexico)

    2010-01-15

    A method is presented for estimating the initial temperature in geothermal-reservoir formations. The method is based on control theory where the measured temperatures or temperature logs are compared with corresponding simulated temperatures for different times with the well closed. The comparison is made using a control algorithm that makes changes to the originally assumed reservoir temperatures and performs iterations until the best fit between the temperature logs and the simulated temperatures is obtained. The simulation of fluid transport and heat in the well includes the processes of circulation and stop in the presence of circulation losses, modeled on macroscopic balances of momentum and energy. The transport processes in the formation regard the reservoir as an isotropic porous medium and fluid flow is described by Darcy's law. This model generates the fields of temperatures, pressures and speeds as a function of time and space. The method was tested with data from well LV-3 in Las Tres Virgenes geothermal field, Baja California Sur, Mexico. The estimated temperatures of the undisturbed formation-or initial temperatures-are compared within {+-}15 degrees Celsius with the measured temperatures, which is an acceptable outcome from an engineering point of view. [Spanish] Se presenta un metodo para la estimacion de la temperatura inicial en las formaciones de yacimientos geotermicos. El metodo se basa en la teoria de control donde las temperaturas medidas o registros de temperatura se comparan con las correspondientes temperaturas simuladas a diferentes tiempos con el pozo cerrado. La comparacion se hace usando un algoritmo de control el cual hace cambios a las temperaturas de yacimiento originalmente supuestas y realiza iteraciones hasta que se obtiene el mejor ajuste entre los registros de temperatura y las temperaturas simuladas. La simulacion del transporte de fluidos y calor en el pozo incluye los procesos de circulacion y paro en presencia de

  18. Development of an analytic method for arsenic's determination in lime and tortilla; Desarrollo de un metodo analitico para determinacion de arsenico en cal y tortilla

    Energy Technology Data Exchange (ETDEWEB)

    Huato Soberanis, Julio; Ogura, Tetsuya [Universidad Autonoma de Guadalajara, Guadalajara, Jalisco (Mexico)

    1995-02-01

    A spectrophotometric method to determine As in tortilla and lime has been optimized, modifying the AsH{sub 3} generator. The reaction between arsin (AsH{sub 3}){sub 4} and diethyldithiocarbamate of Ag (AgDDC); was followed spectrophotometrically. The conditions under which the As remains in the ash during the calcination of the tortillas were studied. It was found that when they were heated in a quartz tube with a careful control of the air flow and oxygen, as well as the heating temperature, the arsenic loss in minimized. [Spanish] Se ha optimizado el metodo para determinar As en la tortilla y cal mediante espectrometria en el visible del color producido en la reaccion entre Arsina (AsH{sub 3}){sub 4} y dietilditiocarbamato de plata (AgDDC); modificando el generador de AsH{sub 3}. Se han buscado las condiciones en las que el arsenico permanece en las cenizas de la calcinacion de las tortillas; encontrandose que las tortillas deben calentarse en un tubo de cuarzo con control del flujo de aire y oxigeno asi como de la temperatura de calentamiento.

  19. Control Methods Used in the Department of Metallurgy for Structure and Fuel Elements; Methodes de Controle Utilisees au Departement de Metallurgie pour les Elements de Structure et les Elements Combustibles; Metody kontrolya struktury toplivnykh ehlementov v departamente metallurgii; Metodos de Control Utilizados en el Departamento de Metalurgia para los Elementos Estructurales y Combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Destribats, Marie-Therese; Allain, C.; Prot, A.; Thome, P. [Centre d' Etudes Nucleaires de Saclay (France)

    1965-09-15

    reactores indujeron a utilizar y a perfeccionar numersos metodos no destructivos destinados a inspeccionar los distintos materiales que intervienen en la construccion de esos reactores; en particular radiografia y gamma- grafia, metodos ultrasonicos y empleo de corrientes de Foucault.. A continuacion se enumeran las operaciones de control llevadas a cabo durante la construccion de reactores pertenecientes a las familias EdF (grafito-gas) y EL 4 (agua pesada), y de elementos colaminados; se hace hincapie en ciertos aspectos caracteristicos de estos metodos, algunos de los cuales son ya bien conocidos. Familia EdF: metodos ultrasonicos para medir el espesor de las paredes de tubos de uranio o aleacion de uranio ; localizacion de las cavidades en esos tubos por gammagraffa; empleo de medios ultrasonicos para control de los tratamientos termicos a que se someten los tubos; empleo de procedimientos ultrasonicos para buscar fallas (inclusiones, grietas) en las palanquillas y barras con que se elaboran las vainas de Mg-Zr; control de la estanqueidad de los elementos mediante exudacion de helio. EL4: uso de metodos ultrasonicos y corrientes de Foucault para medir espesores de pared en tubos de Zircaloy, sean de fuerza o de guia; empleo de metodos ultrasonicos para inspeccionar tubos de fuerza de Zircaloy y vainas de acero inoxidable; radiografia al vacio de vainas de Be; control de la estanqueidad de barras huecas mediante exudacion de helio. Elementos colimados: medida del espesor de vainas mediante el empleo de corrientes de Foucault pulsadas; verificacion de la posicion del alma en tubos y placas mediante radiografia, recuento gamma y aplicacion de corrientes de Foucault pulsadas; control de la homogeneidad del combustible por recuento gamma; deteccion de defectos en barras de Zr-U mediante procedimientos ultrasonicos y gammagraficos; determinacion de las zonas despegadas en placas, empleando metodos ultrasonicos, corrientes de Foucault pulsadas y medicion de resistividad. Estos

  20. Validation of Hiriart equation to compute steam production by the lip pressure method; Validacion de la ecuacion de Hiriart para calculo de gasto de vapor por el metodo de presion de labio

    Energy Technology Data Exchange (ETDEWEB)

    Flores Armenta, Magaly [Gerencia de Proyectos Geotermoelectricos de la Comision Federal de Electricidad, Morelia (Mexico)

    1996-09-01

    Mainly in new geothermal wells, it is necessary to evaluate the production in a very fast, simple and not expensive way, to know the convenience to install surface equipment, such as silencers and separators, to drive the steam to the commercial gathering system. In practice, one of the most known methods is the lip pressure one, which requires a simple set of installations. The objective of this paper is to validate the steam flow rate calculated by the lip pressure method, with respect to the ASME method. The ASME method is known for its accuracy, and is done by measuring the steam and liquid after a high pressure separator, by an orifice plate of known diameter and a triangular weir. Results of the validation show up the feasibility of application of the lip pressure method by using a simple adjustment equation. Percentage of mistake results less than 1%, without any notable influence of the production enthalpy. That equation to be applied in a general case, is as follows: Q{nu} =(20642)(F*P*D{sup 2}/{radical}h-2000). For the particular case of the Los Azufres geothermal field, the equation is: Q{nu}= 810*P*D{sup 2} [Espanol] En los pozos geotermicos, principalmente en los nuevos, es necesario evaluar su produccion de manera rapida, sencilla y economica, para determinar la conveniencia de instalar equipo superficial, como separadores, silenciadores, etc., que permita la integracion del vapor al sistema comercial de generacion electrica. Para fines practicos uno de los metodos mas conocidos es el de presion de labio, que solo requiere un arreglo sencillo de instalaciones superficiales. En este documento se validan y ajustan los calculos de produccion de vapor por ese metodo de presion de labio, con respecto a las mediciones exactas efectuadas con el metodo ASME. Este ultimo es reconocido internacionalmente por su precision, y se lleva a cabo separando la mezcla obtenida en superficie en un recipiente a presion para medir el vapor a traves de una placa de orificio

  1. Modificação do metodo "kindling" para obtenção de status epilepticus experimental em ratos

    Directory of Open Access Journals (Sweden)

    Carlos J. Reis de Campos

    1980-03-01

    Full Text Available Foi utilizada em nova espécie animal (ratos, uma modificação do método "kindling", introduzida por Taber e col. (1977 para obtenção de status epilepticus experimental. Para isso foram implantados mediante cirurgia estereotáxica, eletrodos duplos, torcidos no hipocampo dorsal de 12 ratos machos albinos. Esses animais foram submetidos, após uma semana de pós-operatório, a 1 segundo de estimulação elétrica de baixa intensidade em forma intermitente, um estímulo por minuto durante 2 horas, desenvolvendo-se em prazo de 30 minutos um estado de epilepsia eletrográfica e comportamental duradoura. Vários padrões de descargas epilépticas eletrográficas foram observados bem como manifestações convulsivas tônico-clônicas. Os animais que foram submetidos a novas sessões de estimulação após 7 e 14 dias mostraram aumento de atividade epiléptica demonstrando uma modificação plástica do hipocampo do rato submetido a estimulação elétrica a qual perdura no tempo. O método permite a obtenção de "kindling" em tempo bem mais curto (horas, comparativamente às técnicas anteriormente descritas (dias, tornando-se um promissor modelo de epilepsia para testes de drogas anticonvulsivantes e para o estudo dos mecanismos fisiopatológicos e bioquímicos envolvidos na descarga epiléptica.

  2. Proposta de metodo de gerenciamento de processos administrativos para organizações prestadoras de serviços

    OpenAIRE

    Benigno Roberto Zaki

    2009-01-01

    Resumo: Desde a metade do século XX e neste início do século XXI, as empresas buscam aumentar a produtividade e melhorar a qualidade de seus produtos e serviços. Para uma organização, muito mais do que a liderança no seu mercado de atuação é a sua capacidade de se manter competitiva. A busca da vantagem competitiva perante o mercado está em satisfazer as expectativas do cliente e implementar com êxito uma série de exigências em termos de eficiência no processo produtivo. O desenvolvimento des...

  3. A new method for the coefficients of the characteristic polynomial; Un metodo nuevo para los coeficientes del polinomio caracteristico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, H.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2006-07-01

    In linear algebra, one can associate an equation to each square matrix: its characteristic equation or secular equation. Starting from this equation, the one characteristic polynomial that codes several important properties of the matrix is obtained: its own values, it determinant and it appearance. The first method to calculate those coefficients of this polynomial were proposed by the french astronomer Urbain Jean Joseph Le Verrier (1811-1877), from then on, many methods have intended to calculate these coefficients. In this work the author proposes a new one method and a bibliographical citation is given where the calculations with others methods that know each other for it, taking like reference the matrix used by Le Verrier are explained. It was concluded that it here proposed, besides being the only mexican method that is knew, has the advantage of being very easy of understanding and of calculating well, in the operations that it carries out, it doesn't use the division and it avoids fractions in matrices whose entrances are whole. This has a great importance for their use in the classroom for their great didactic value and in nuclear reactors and Genetic Engineering. (Author)

  4. A new method for the coefficients of the characteristic polynomial; Un metodo nuevo para los coeficientes del polinomio caracteristico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, H E [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2006-07-01

    In linear algebra, one can associate an equation to each square matrix: its characteristic equation or secular equation. Starting from this equation, the one characteristic polynomial that codes several important properties of the matrix is obtained: its own values, it determinant and it appearance. The first method to calculate those coefficients of this polynomial were proposed by the french astronomer Urbain Jean Joseph Le Verrier (1811-1877), from then on, many methods have intended to calculate these coefficients. In this work the author proposes a new one method and a bibliographical citation is given where the calculations with others methods that know each other for it, taking like reference the matrix used by Le Verrier are explained. It was concluded that it here proposed, besides being the only mexican method that is knew, has the advantage of being very easy of understanding and of calculating well, in the operations that it carries out, it doesn't use the division and it avoids fractions in matrices whose entrances are whole. This has a great importance for their use in the classroom for their great didactic value and in nuclear reactors and Genetic Engineering. (Author)

  5. A review of calculation methods for fast and intermediate reactors; Expose des methodes pour le calcul de reacteurs a neutrons rapides et intermediaires; Obzor metodov rascheta reaktorov na promezhutochnykh i bystrykh nejtronakh; Estudio panoramico de los metodos de calculo de los reactores rapidos e intermedios

    Energy Technology Data Exchange (ETDEWEB)

    Marchuk, G I [Akademiya Nauk, Moskva, Union of Soviet Socialist Republics (Russian Federation)

    1962-03-15

    This paper discusses the development of methods for calculating intermediate and fast reactors. It deals with various approaches to the problems of physical calculation. The calculation of resonance effects is discussed. Consideration is given to multi-group systems of fundamental and conjugate equations, various applications of perturbation theory to the problems of physical reactor calculation, and numerical methods of solving fundamental and conjugate reactor equations, which approximate the method of spherical harmonics. The paper describes an application of the response method to the solution of critical-mass problems, and methods of calculating reactors with hydrogeneous moderators. The fundamental features of an effective one-group reactor model are described. (author) [French] L'auteur examine la mise au point de methodes pour le calcul de reacteurs a neutrons rapides et intermediaires . Il decrit diverses manieres d'aborder les problemes des calculs sur la physique des reacteurs, notamment le calcul des effets de resonance. Il s'attache particulierement aux points suivants: systemes d'equations fondamentales et conjuguees a plusieurs groupes; diverses applications de la theorie des perturbations aux problemes de calculs sur la physique des reacteurs; methodes numeriques pour resoudre les equations fondamentales et conjuguees, voisines de la methode des harmoniques spheriques. L'auteur decrit ensuite une maniere d'appliquer la methode de la reponse aux problemes de la masse critique ainsi que des methodes pour le calcul de reacteurs ralentis a l'hydrogene. Il decrit les caracteristique s fondamentale s d'un modele de reacteur a un groupe effectif. (author) [Spanish] El autor analiza el desarrollo de los metodos de calculo de los reactores nucleares que trabajan con neutrones rapidos y con neutrones intermedios. Examina diversos planteos de los problemas del calculo fisico. Indica la forma de tomar en cuenta los efectos de resonancia y menciona los sistemas

  6. The Role of Non-Destructive Testing in the Los Alamos Reactor Programme; Role des Essais Non Destructifs dans le Programme de Reacteurs de los Alamos; Rol' nedestruktivnykh ispytanij materialov v Los-Alamosskoj reaktornoj programme; Papel de los Metodos de Ensayo No Destructivo en el Programa de Reactores de Los Alamos

    Energy Technology Data Exchange (ETDEWEB)

    Tenney, G. H. [University of California, Los Alamos Scientific Laboratory, Los Alamos, NM (United States)

    1965-10-15

    temperature UHTREX, actuellement en construction, on a etudie par microiadiogiaphie et au moyen de microscopes electroniques des grains de carbure d'uranium enrobes de carbone pyrolytique, d'un diametre de 150 {mu}m, pour evaluer la translocation de l'uranium en fonction de la temperature. On determine la quantite et l'uniformite de la charge d'uranium dans les elements au graphite d'UHTREX au moyen de compteurs a scintillation specialement concus. Environ 90% des travaux effectues a ce sujet n'ont encore fait l'objet d'aucune publication. (author) [Spanish] El Laboratorio Cientifico de Los Alamos, explotado por la Universidad de California por encargo de la Comision de Energia Atomica de los Estados Unidos, viene ocupandose desde hace mas de veinte afios del proyecto, diseno y construccion de reactores nucleares de cuatro tipos generales; a saber, de investigacion, de potencia, de propulsion espacial y para conjuntos criticos. El llamado Grupo de ensayos no destructivos colabora practicamente en todas las actividades y proyectos del laboratorio. En la presente memoria se exponen algunos de los metodos de ensayo no destructivo y sus aplicaciones, establecidos para uso en el programa de reactores. El programa LAPRE (Los Alamos Power Reactor Experiment) se basa en el empleo de una solucion de fosfato de uranio a alta temperatura. La solucion es muy corrosiva y todas las piezas que entren en contacto con ella deben ir revestidas de oro. Durante el proceso de produccion de chapa de oro laminada a partir de lingotes, se han utilizado procedimientos radiograficos especiales para inspeccionar el metal. Las juntas soldadas se examinaron del mismo modo, y ademas se establecio un metodo para comprobar la presencia de impurezas incrustadas en la superficie de la chapa de oro. El concepto fundamental en que se basa el programa LAMPRE (Los Alamos Molten Plutonium Reactor Experiment) es la utilizacion como combustible de plutonio metalico liquido en vez de solido. El combustible esta

  7. The Comparative Accuracy of the 4 {pi} Liquid Scintillation Counting Method of Radioisotope Standardization; L'exactitude comparee de la methode de comptage 4 {pi} a scintillateurs liquides pour l'etalonnage des radioisotopes; Sravnitel'naya tochnost' 4 {pi} zhidkogo stsintillatsionnogo metoda podscheta standartiziruemykh radioizotopov; Exactitud del metodo de recuento con centelleador liquido 4 {pi} para normalizar radioisotopos, comparada con la de otros metodos

    Energy Technology Data Exchange (ETDEWEB)

    Steyn, J [National Physical Research Laboratory, Pretoria (South Africa)

    1960-06-15

    The accuracy of the 4 {pi} liquid scintillation counting method of standardizing {beta} emitters was compared to 4 l{pi} {beta}-{gamma} coincidence counting for the nuclides Co{sup 60}, I{sup 131} and Au{sup 198}. For P{sup 32} the liquid counting results were compared to 4 {pi} proportional counting. The efficiency of the liquid scintillation counting method was found to be energy dependent, dropping to about 97.5% for Co{sup 60} which was the lowest energy {beta} emitter investigated. (author) [French] La precision de la methode de comptage 4 {pi} a scintillateurs liquides pour l'etalonnage des emetteurs {beta} a ete comparee au comptage par coincidences 4 {pi} {beta}-{gamma} pour le So{sup 60}, le I{sup 131} et le Au{sup 198}. Dans le cas du P{sup 32}, les resultats du comptage au liquide ont ete compares a ceux du comptage 4 {pi} proportionnel. On a constate que le rendement de la methode de comptage a scintillateurs liquides variait en fonction de l'energie emise et qu'il descendait a environ 97.5% pour le Co{sup 60} qui, de tous les emetteurs {beta} etudies, emet l'energie la plus faible. (author) [Spanish] El autor compara la precision del metodo de recuento con centelleador iquido 4 {pi} para normalizar emisores {beta} con la del metodo de coincidencias {beta}-{gamma} 4 {pi}, para los siguientes nuclidos: So{sup 60}, I{sup 131} y Au{sup 198}. En el caso del P{sup 32}, confronta los resultados del primer metodo con los obtenidos mediante el recuento proporcional 4 {pi}. Comprueba que la eficacia del metodo de recuento con centelleador liquido depende de la energia y desciende al 97.5%, aproximadamente, para el Co{sup 60}, que fue el emisor {beta} mas debil que se investigo. (author) [Russian] Tochnost' 4 {pi} zhidkogo stsintillyatsionnog o metoda podscheta standartiziruemogo {beta}-izluchatelya sravnivalas' s 4 {pi} {beta}-{gamma} metodom podscheta na sovpadeniyakh dlya izotopov So{sup 60}, I{sup 131} i Au{sup 198}. Dlya R{sup 32} rezultaty zhidkogo

  8. 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

  9. Non-Destructive Methods for Determining Burn-Up in Nuclear Fuel; Methodes Non Destructives d'Evaluation du Taux de Combustion dans le Combustible Nucleaire; Metody opredeleniya vygoraniya v yadernom toplive bez razrusheniya obraztsa; Metodos No Destructivos para Determinai el Grado de Combustion de los Elementos Combustibles Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    McGonnagle, W. J. [Illinois Institute of Technology, Chicago, IL (United States)

    1966-02-15

    gamma emis par les matieres radioactives qui se forment au cours du processus de fission, en particulier le spectrometre a cristal courbe, le spectrometre Compton magnetique, le spectrometre Compton a coiencidence et le spectrometre a scintillation. Parmi les autres methodes non destructives, on peut citer l'activation de feuilles, la transmission des neutrons, l'analyse par activation, la mesure des rayonnements gamma de capture et la mesure des neutrons instantanes et differes. Le memoire etudie les principes essentiels des differents appareils et procedes ci-dessus, leur precision et leurs limitations. Le memoire presente des methodes non destructives utilisant les isotopes stables produits au cours du processus de fission. Dans les mesures au moyen de ces isotopes, le schema d'irradiation n'a qu'une importance secondaire et le temps de refroidissement ne joue aucun role. De plus, on dispose de donnees nucleaires plus precises sur les produits stables de la fission. Il semble que les plus utiles de tous les isotopes stables produits au cours du processus de fission soient ceux du zirconium, du molybdene, du ruthenium et du neodyme. Le memoire etudie l'interet de ces methodes d'analyse non destructive. (author) [Spanish] Los metodos no destructivos son de gran utilidad para medir cuantitativamente el grado de combustion. El metodo ideal seria el que no requiriese datos especiales sobre los espectros neutrtfni- cos, las irradiaciones precedentes o el tiempo de enfriamiento. Los elementos combustibles irradiados llevan en cierto modo el registro de su grado de combustion. Este registro consiste en el conjunto de los isotopos radiactivos y estables resultantes del proceso de fision. Desgraciadamente, tanto con los metodos no destructivos como con los destructivos el espectro neutronico, las irradiaciones precedentes y el tiempo de enfriamiento influyen sobre dicho registro. Analogamente, la falta de datos nucleares precisos tales como las secciones eficaces nucleares

  10. The Control of Fast Reactors: Current Methods and Future Prospects; Controle des Reacteurs a Neutrons Rapides. Methodes Actuelles et Perspectives d'Avenir; Upravlenie reaktorami na bystrykh nejtronakh. sushchestvuyushchie metody i dal'nejshie perspektivy; Control de Reactores Rapidos: Metodos Actuales y Perspectivas

    Energy Technology Data Exchange (ETDEWEB)

    Loewenstein, W. B. [Argonne National Laboratory, IL (United States)

    1964-06-15

    examina los actuales mecanismos de control de los reactores de neutrones rapidos. En la medida de lo posible aprovechan el control por medio de fugas neutronicas. Si este metodo no es aplicable se suele recurrir al control por desplazamiento de los materiales del cuerpo. Igualmente se recurre a un control limitado utilizando un absorbente. Ninguno de estos metodos presenta ventajas considerables cuando se aplica a los'grandes reactores de potencia regeneradores, a menos que la razon interna de regeneracion (del cuerpo) sea muy elevada. Se requerira mucha habilidad para utilizar metodos de control basados en el empleo de un absorbente o en el desplazamiento espectral sin afectar en grado apreciable la economia neutronica deseada y a menudo necesaria. El autor cita algunos resultados preliminares obtenidos con sistemas perfeccionados. La reactividad de control viene determinada por los requisitos relativos a la parada del reactor, al ciclo del combustible (exceso de reactividad) y, en menor medida, por la realimentacion dominante. Pueden especificarse perfectamente los requisitos relativos al exceso de reactividad para un ciclo de combustible determinado, pero esos requisitos varian de modo considerable en otros sistemas similares que funcionen con distintos ciclos de combustible. A partir de ciertos limites, pueden fijarse casi arbitrariamente les requisitos relativos al exceso de reactividad. Sin embargo, existen algunas consideraciones generales que rigen la determinacion de este parametro. Se tienen en cuenta dichas consideraciones al examinar la reactividad de control en los actuales reactores de neutrones rapidos comparandola a la cantidad realmente necesaria para el funcionamiento de reactores de potencia regeneradores y neutrones rapidos. El autor cita parametros tipicos de potencia y de realimentacion en funcion de la temperatura a fin de determinar su influencia en los requisitos relativos a la reactividad de control. Los metodos utilizados para predecir la

  11. Double-Sampling Method for Carrying Out Quality Control of a Fabrication Process; Methode du Double Echantillonnage pour le Controle de la Qualite d'un Procede de Fabrication; Metod dvukh obraztsov dlya osushchestvleniya kontrolya za kachestvom v protsesse izgotovleniya; Metodo de Muestreo Doble para el Control de Calidad de un Proceso de Fabricacion

    Energy Technology Data Exchange (ETDEWEB)

    Cerrolaza, J. A.; Lago, A.; Montojo, Rosa M. [Junta de Energia Nuclear, Madrid (Spain)

    1966-02-15

    taille des echantillons et le degre de controle sont determines en fonction des risques inherents a la premiere et a la deuxieme condition. Bien que cette methode soit applicable, elle est fort peu pratique quand il s'agit de controler des pieces de reacteurs, car les echantillons qu'elle exige sont de dimensions excessives. Le memoire expose une methode analogue a celle de Cave mais qui consiste a etablir un double echantillonnage dans lequel la taille moyenne de l'echantillon est beaucoup plus reduite. (author) [Spanish] La fabricacion de componentes que han de ser empleados en reactores nucleares presenta dos caracteristicas fundamentales que condicionan el proceso: la exigencia de una calidad muy elevada que limita a valores muy bajos el porcentaje de piezas defectuosas admisibles, y el coste muy alto de cada uno de los componentes. Estas dos condiciones obligan a que el control durante la fabricacion posea una curva de operacion con potencia muy elevada, y, por otra parte, a que el numero de piezas destruidas en cada inspeccion sea un pequeno como se pueda, siempre que se cumpla la condicion anterior. Los metodos usuales de control, basados en fijar el riesgo de primera especie, no son aplicables, ya que en general su eficacia no es suficiente. Cave ha desarrollado un metodo en que, tanto el tamano de las muestras como los limites de control, se fijan en funcion de los riesgos de primera y segunda especie. Este metodo, si bien es aplicable, resulta poco practico en el control de componentes para reactores, ya que se necesitan muestras de un tamano excesivamente grande. En el presente trabajo se desarrolla un metodo semejante al de Cave, pero en el que se establece un muestreo doble, con lo que el tamano medio de la muestra es mucho mas reducido. (author) [Russian] Pri izgotovlenii komponentov dlja ispol'zovanija v reaktorah reshajushhee znachenie imejut dva osnovnyh faktora, a imenno: neobhodimost' obespechit' ochen' vysokoe kachestvo, dopuskajushhee lish' ochen

  12. Development of a chromatographic method for the study of the stability and compatibility of Mexican fuel oils; Desarrollo de un metodo cromatografico para el estudio de estabilidad y compatibilidad de combustoleos mexicanos

    Energy Technology Data Exchange (ETDEWEB)

    Blass Amador, Georgina; Panama Tirado, Luz Angelica [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1992-11-01

    compatibility of fuel oil mixes. [Espanol] En Mexico, la mayoria de la energia electrica producida proviene del uso de combustibles residuales pesados conocidos como combustoleos los cuales han sufrido disminuciones en la calidad debido a una combinacion de factores, entre los que destaca el de los cambios en el proceso de refinacion. Es necesario desarrollar metodos que sean capaces de indicar la inestabilidad (formacion de sedimento o incremento en viscosidad durante el almacenamiento o calentamiento) o incompatibilidad (formacion de sedimento al mezclar dos o mas) de los combustoleos utilizados en las centrales termoelectricas. El objetivo de este trabajo fue el desarrollar una prueba alternativa para el estudio de la compatibilidad y/o estabilidad de combustoleos mexicanos empleando cromatografia de liquidos de alta resolucion (CLAR) y asi poder determinar aspectos estructurales del combustoleo que determinan su estabilidad. Dado que la formacion de sedimentos ocurre cuando el poder disolvente del combustible es inadecuado para mantener los asfaltenos en solucion, es importante conocer la medida del poder disolvente o aromaticidad del diluyente; asi pues, la primera parte de este trabajo se centro en la determinacion del perfil de compuestos aromaticos de los diluyentes de los combustoleos, la otra parte se dedico a la determinacion del perfil de distribucion de los pesos moleculares de los asfaltenos presentes en los combustoleos. Los perfiles de la fraccion aromatica, asi como los de distribucion de pesos moleculares se determinaron empleando cromatografia de liquidos, en la que se empleo una variedad de columnas y de disolventes. Se efectuo una combinacion de pruebas de rutina tales como contenido de asfaltenos, equivalencia de tolueno, viscosidad, etcetera con el fin de obtener correlaciones con el metodo cromatografico desarrollado. En este articulo se discute solo la seccion correspondiente a la obtencion del perfil de contenidos de aromaticos de los combustoleos. Se

  13. Rapid Methods of Determining Internal Radioactive Contamination; Methodes Rapides Permettant d'Evaluer la Contamination Radioactive Interne; 0411 042b 0414 ; Metodos Rapidos para Determinar la Contaminacion Radiactiva Interna

    Energy Technology Data Exchange (ETDEWEB)

    Sedlet, J.; Fairman, W. D.; Robinson, J. J. [Industrial Hygiene and Safety Division, Argonne National Laboratory, Argonne, IL (United States)

    1965-06-15

    , il est possible de l'atteindre. Les emetteurs alpha et les emetteurs beta purs de faible energie presentent la plus grande difficulte du fait que les charges corporelles de ces nucleides doivent ordinairement etre evaluees par une analyse radiochimique du sang ou des excreta. Les auteurs decrivent les separations radiochimiques qui ont ete mises au point et appliquees au Laboratoire national d*Argonne pour doser rapidement ces emetteurs. Le memoire examine d'autre part certains problemes que pose l'obtention de charges corporelles a partir des taux d'excretion, peu apres l'absorption. Pendant ce laps de temps, l'excretion se modifie rapidement et, pour beaucoup de nucleides importants en l'occurrence, a un taux inconnu. Pour les taux d'absorption tres eleves ou tres bas, il n'est pas necessaire de connaitre immediatement la mesure precise des charges corporelles. Cependant, il existe pour chaque radionucleide une dose d'absorption intermediaire pour laquelle il faut pouvoir -determiner sans trop tarder la charge corporelle avec une precision relative, ce qui permet de decider ou non le traitement. Les auteurs examinent les proprietes et les facteurs qui rendent le dosage des elements osteotropes particulierement difficile dans les cas d'accident. (author) [Spanish] Los metodos aplicables en casos de accidente o en situaciones excepcionales tienen que ser suficientemente rapidos y sensibles para proporcionar resultados en plazos que faciliten el tratamiento eficaz de las personas sobreexpuestas. Los plazos disponibles dependeran, en parte, del numero de individuos irradiados. Si se logra obtener en pocas horas resultados utiles, se habra cumplido la funcion primaria de las mediciones rapidas de la carga corporal, que es la de contribuir en toda la medida de lo posible a reducir los danos debidos .a radionuclidos depositados internamente. Los metodos que se citan a continuacion pueden servir para evaluar la contaminacion interna en casos de accidente: recuento gamma y

  14. A Comparison of Radioisotope Methods for River Flow Measurement; Comparaison de methodes radioisotopiques de mesure du debit des cours d'eau; Sravnenie radioizotopnykh metodov izmereniya rechnykh stokov; Comparacion de los metodos radioisotopicos para medir el caudal de los rios

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, C. G.; Smith, D. B. [Wantage Research Laboratory Atomic Energy Research Establishment Wantage, Berks (United Kingdom)

    1963-08-15

    consequent d'aucune utilite comme indicateur dans la mesure du debit d'un cours d'eau s'il n'est pas employe avec un entraineur; mais avec un entraineur, on observe une amelioration marquee de l'exactitude des resultats. {sup 24}Na et {sup 82}Br n'ont accuse aucune adsorption a des distances bien superieures a celles qui sont necessaires pour assurer une brassage lateral; dans la riviere la plus lente, toutefois, on a note, a plus de 660 m du point d'injection, une certaine diminution de {sup 24}Na. Le memoire contient quelques observations sur la dispersion laterale et longitudinale qui, a une certain degre, influe sur l'application generale des methodes radioisotopiques a la mesure du debit des cours d'eau. Les auteurs parviennent a la conclusion que les trois methodes donnent des resultats satisfaisants. Ils preferent, en fin'de compte, la methode d'echantillonnage continue, car elle donne les resultats les plus exacts pour, la quantite minimum d'indicateur. (author) [Spanish] Los autores han empleado los procedimientos de dilucion, muestreo continuo y recuento total en un estudio comparativo de los metodos radioisotopicos de medicion de caudales fluviales en el arroyo Aylburton, en Gloucestershire, y en los rios Usway Burn y Alwin, en Northumberland. Se trata de tres tios de caracteristicas geologicas diferentes, cuyos caudales oscilan entre 2,5 1/s y 3 m{sup 3}/s. En todos los metodos de medicion de caudales que emplean indicadores, la distancia entre el punto en que se efectua la medicion y el punto en que se anade el indicador tiene que ser suficientemente grande para que se produzca una mezcla lateral completa. Por otra parte, no debe ser excesiva para evitar que la dipersion longitudinal supere cierto valor o que se pierda parte del indicador por adsorcion en el lecho fluvial. La dispersion depende de las caracteristicas hidraulicad de la corriente y es inherente al metodo. Por su parte, la adsorcion depende del indicador elegido y de las caracteristicas geologicas

  15. Protocolo Nacional para la Evaluacion de Disturbios en Suelos Forestales; Volumen II: Metodos complementarios, estadística y recoleccion de datos

    Science.gov (United States)

    Deborah S. Page-Dumroese; Ann M. Abbott; Thomas M. Rice

    2013-01-01

    Este documento-El Volumen II: Métodos complementarios, estadística y recolección de datos- define las bases, los métodos estadísticos y de almacenamiento de datos de un Protocolo Nacional para la Evaluación de Disturbios en Suelos Forestales. Esta guía técnica proporciona las bases de un método consistente, con definiciones comunes, para generar datos de alta calidad,...

  16. Modelling of pressurized water reactor fuel, rod time dependent radial heat flow with boundary element method; Modeliranje spremenljivega radijalnega toplotnega toka tlacnovodne gorivne palice z metodo robnih elementov

    Energy Technology Data Exchange (ETDEWEB)

    Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    The basic principles of the boundary element method numerical treatment of the radial flow heat diffusion equation are presented. The algorithm copes the time dependent Dirichlet and Neumann boundary conditions, temperature dependent material properties and regions from different materials in thermal contact. It is verified on the several analytically obtained test cases. The developed method is used for the modelling of unsteady radial heat flow in pressurized water reactor fuel rod. (author)

  17. Empleo de metodos numericos para el ajuste de los coeficientes de difusividad (D) y convectivo de transferencia de masa (hm) en el secado de alimentos

    OpenAIRE

    Arranz Saiz, Francisco Javier; Correa Hernando, Eva Cristina; Jiménez Ariza, Heidi Tatiana; Diezma Iglesias, Belen; Garcia-Herrero, Javier; Robla Villalba, José Ignacio; Barreiro Elorza, Pilar

    2011-01-01

    Los modelos matemáticos de transferencia de humedad desde el alimento al medio circundante durante el proceso de secado dependen de dos parámetros: coeficiente de difusividad efectiva D y coeficiente convectivo de transferencia de masa hm, cuya determinación experimental se basa en la consideración de un sistema difusivo sencillo de solución analítica bien conocida. Para el caso, que aquí nos ocupa, de difusión monodimensional en una lámina infinita de grosor 2l, la solución analítica de ...

  18. Methods for the control of NOx and particles in the combustion; Metodos para el control de NOx y particulas en la combustion

    Energy Technology Data Exchange (ETDEWEB)

    Romo Millares, Cesar A. [Instituto de Investigaciones Electrica, Cuernavaca (Mexico)

    1996-12-31

    This present the techniques and equipment of control of transmissions for thermoelectric power stations appear that have mayor possibilities of being considered in the future immediate within the national energetic panorama and the frame established by the environmental normative. The subject polluting compounds to overhaul are oxides of nonburned nitrogen and particles [Espanol] Se presentan las tecnicas y equipos de control de emisiones para centrales termoelectricas que tienen mayores posibilidades de ser consideradas en el futuro inmediato dentro del panorama energetico nacional y el marco establecido por la normatividad ambiental. Los compuestos contaminantes sujetos a revision son los oxidos de nitrogeno y las particulas inquemadas

  19. Methods for the control of NOx and particles in the combustion; Metodos para el control de NOx y particulas en la combustion

    Energy Technology Data Exchange (ETDEWEB)

    Romo Millares, Cesar A [Instituto de Investigaciones Electrica, Cuernavaca (Mexico)

    1997-12-31

    This present the techniques and equipment of control of transmissions for thermoelectric power stations appear that have mayor possibilities of being considered in the future immediate within the national energetic panorama and the frame established by the environmental normative. The subject polluting compounds to overhaul are oxides of nonburned nitrogen and particles [Espanol] Se presentan las tecnicas y equipos de control de emisiones para centrales termoelectricas que tienen mayores posibilidades de ser consideradas en el futuro inmediato dentro del panorama energetico nacional y el marco establecido por la normatividad ambiental. Los compuestos contaminantes sujetos a revision son los oxidos de nitrogeno y las particulas inquemadas

  20. 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.

  1. Metodo integrado para la gestion de Universidades basado en el Balanced Scorecard (bsc y el modelo europeo de calidad (efqm: caso U.C.S.M.

    Directory of Open Access Journals (Sweden)

    Edwing Jesús Ticse Villanueva

    2010-06-01

    Full Text Available http://dx.doi.org/10.5007/1983-4535.2010v3n1p01   El presente Trabajo realiza la propuesta de un Método Integrado  que permite mejorar la competitividad en la Gestión de las Universidades, este método se basa: En una herramienta de Gestión Estratégica: el Balanced Scorecard (BSC, que fue creada por Kaplan y Norton en 1992; y En el Modelo  Europeo de Calidad (EFQM, que fue desarrollado como un Modelo de Excelencia en 1991. La metodología utilizada consiste en analizar las características, ventajas y limitaciones del BSC y el EFQM para aplicarlos simultáneamente en la gestión de Tomando como base los 9 criterios del EFQM, se desarrolla un Mapa Estratégico del BSC, para que se alineen todas las perspectivas hacia el logro de los Objetivos Estratégicos de la Organización. El método integrado planteado se aplica en el caso de la Universidad Católica de Santa María (Arequipa- Perú, donde se analiza las ventajas que implicaría la implementación del mismo

  2. Metodo integrado para la gestion de Universidades basado en el Balanced Scorecard (bsc y el modelo europeo de calidad (efqm: caso U.C.S.M.

    Directory of Open Access Journals (Sweden)

    Horacio Vicente Barreda Tamayo

    2010-12-01

    Full Text Available El presente Trabajo realiza la propuesta de un Método Integrado que permite mejorar la competitividad en la Gestión de las Universidades, este método se basa: En una herramienta de Gestión Estratégica: el Balanced Scorecard (BSC, que fue creada por Kaplan y Norton en 1992; y En el Modelo Europeo de Calidad (EFQM, que fue desarrollado como un Modelo de Excelencia en 1991. La metodología utilizada consiste en analizar las características, ventajas y limitaciones del BSC y el EFQM para aplicarlos simultáneamente en la gestión de Tomando como base los 9 criterios del EFQM, se desarrolla un Mapa Estratégico del BSC, para que se alineen todas las perspectivas hacia el logro de los Objetivos Estratégicos de la Organización. El método integrado planteado se aplica en el caso de la Universidad Católica de Santa María (Arequipa- Perú, donde se analiza las ventajas que implicaría la implementación del mismo.

  3. 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

  4. Fundamentos del metodo cientifico

    Directory of Open Access Journals (Sweden)

    Badii, M. H.

    2004-01-01

    Full Text Available El objetivo de esta obra no radica en realizar una búsqueda exhaustiva de la literatura en el tema, sino, sentar las bases del método científico, notando los aspectos filosóficos e éticos de la ciencia. Se presentan los conceptos y definiciones fundamentales relacionados con la metodología de la investigación científica. Se maneja el concepto de la toma de los datos válidos como un requisito básico en cualquier trabajo científico. Se pone a disposición del lector un modelo denominado el ECOEE que es una herramienta poderosa para establecer puntos de comparación e discusión entre los resultados de diferentes trabajos científicos. Finalmente, ofrece unas sugerencias de que hacer o no hacer en cuanto a realizar un trabajo de investigación.

  5. 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

  6. Aplicação de metodos de otimização para o calculo do equilibrio quimico e de fases combinados para processos com gas de sintese

    OpenAIRE

    Consuelo Cristina Gomes Silva

    2008-01-01

    Resumo: Essa pesquisa consiste em aplicar métodos de otimização global para o cálculo do equilíbrio químico e de fases combinados para misturas com gás de síntese. O gás de síntese tem grande interesse industrial, pelas inúmeras possibilidades de produção de diversos compostos químicos. Dessa forma, é fundamental conhecer as condições termodinâmicas que favoreçam a obtenção de determinado produto. A aplicação de métodos de otimização global é de grande interesse para a determinação do equilíb...

  7. Comparative Study of the Methods Used for the Computer Resolution of Composite Gamma-Ray Spectra; Etude Comparative des Methodes Utilisees pour la Resolution de Spectres Gamma Complexes au Moyen d'un Ordinateur; Sravnitel'noe izuchenie metodov razresheniya sostavnykh gamma-spektrov pri pomoshchi schetno-reshayushchego ustrojstva; Estudio Comparativo de los Metodos Aplicados para Resolver Espectros Gamma Complejos Mediante Calculadoras

    Energy Technology Data Exchange (ETDEWEB)

    DeHaan, A. Jr.; Leventhal, L.; Benson, P. [Tracerlab, Richmond, CA (United States)

    1965-10-15

    'a resolution; 2. Resolution du pic sans elimination; 3. Etablissement d'un ensemble d'equ'ations lineaires simultanees du meme ordre que le nombre de radionucleides presents dans le melange et solution de ces equations; 4. Evaluation des concentrations non connues par la methode des moindres carres, classique ou ponderee; 5. Combinaison des methodes statistiques et des methodes des moindres carres utilisant la regression lineaire multiple progressive, en s'efforcant d'integrer des processus decisifs dans l'analyse. Pour chaque methode d'analyse appliquee a des melanges de radionucleides, les auteurs fournissent une estimation statistique des erreurs de mesure de la concentration des radionucleides. (author) [Spanish] La resolucion de mezclas complejas de emisores gamma se ha simplificado gracias al empleo de espectrometros gamma muy perfeccionados en asociacion con calculadoras numericas. En la actualidad, los laboratorios pueden en general disponer de estos instrumentos o tener facilmente acceso a ellos. En la memoria se examinan los metodos seguidos en el laboratorio de los autores para determinar la concentracion de cada radionuclido emisor gamma contenido en muestras de precipitaciones radiactivas, y se comparan los resultados obtenidos por los diferentes metodos matematicos aplicables a las muestras compuestas. Los datos proporcionados por las calculadoras se comparan con los resultados obtenidos por analisis radioquimico de la muestra. Se han analizado muestras binarias por los metodos mencionados, operacion que se extendio despues a las mezclas integradas por muchos componentes. Se ha elaborado un metodo de computo que normaliza los espectros gamma para reducir al mfnimo el efecto de la deriva del espectrometro a largo plazo, y representa el espectro segun un sistema de coordenadas, una de las cuales es la energia. Se han investigado los efectos de un 'fotopico imprevisto', de una componente de intensidad cero y de picos superpuestos en la resolucion segun los

  8. Automatic Inspection of Co-Laminated Elements; Controle Automatique d'Elements Colamines; Avtomaticheskij kontrol' ul'trazvukom sovmestno prokatannykh ehlementov; Metodo para la Verificacion Automatica de Elementos Colaminados

    Energy Technology Data Exchange (ETDEWEB)

    Destribats, Marie-Therese [Centre d' Etudes Nucleaires de Saclay (France); Dory, J. [Realisations Ultrasoniques Meaux (S. et M.) (France)

    1965-09-15

    'un disque rotatif. Chaque traducteur est mis en service tous les 1/6 de tour par l'intermediaire de relais commandes par des secteurs aimantes. De la sorte, le pinceau ultrasonore decrit sur la surface de la plaque une serie d'arcs de cercle jointifs. La quantite d'energie transmise a travers la plaque est mesuree avec precision en chaque point. Le resultat de la mesure est applique a un enregistreur special qui donne une reproduction en vraie grandeur de l'element examine (le papier de l'enregistreur est de type ordinaire; l'inscription s'effectue au moyen de papier carbone). Sur l'enregistrement obtenu, les defauts apparaissent sous forme de taches foncees. L'appareillage peut fonctionner, soit en tout ou rien avec un seuil de marquage predetermine, soit en demi-teintes. Les traducteurs, dont la frequence peut varier de 3 a 15 MHz, sont excites par des impulsions breves. Les traducteurs utilisies, en titanate de baryum, ont un diametre de 3 mm. Le pas d'exploration est variable de 0,3 a 0,6 mm, la vitesse lineaire etant de 5 a 20 mm par seconde. L'ensemble peut recevoir des elements ayant les dimensions maximales suivantes: largeur 150 mm, longueur 2000 mm, epaisseur 25 mm. Les essais effectues jusqu'a ce jour ont permis de determiner des zones decollees de 0,5mm diametre dans de elements de 2 mm d'epaisseur. (author) [Spanish] Se describe en esta memoria un aparato automatico destinado a verificar, por un metodo ultrasoniso, la calidad de las vainas fabricadas con elementos colaminados. Estos elementos movidos por rodillos pasan bajo el haz ultrasonico sumergidos en una cuba de acero inoxidable llena de agua. Un reflector auxiliar permite utilizar los mismos transductores para la emision y la recepcion. La exploracion de las placas se lleva a cabo mediante seis transductores dispuestos en la periferia de un disco rotativo. Los transductores entran en funcion uno tras otro, cada 1/6 de vuelta, por intermedio de un sistema reles accionados por sectores imanados. De esta

  9. 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

  10. 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.

  11. 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

  12. The use of genetic algorithms with niching methods in nuclear reactor related problems; A utilizacao dos metodos de nichos dos algoritmos geneticos na otimizacao de problemas de reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner Figueiredo

    2000-03-01

    Genetic Algorithms (GAs) are biologically motivated adaptive systems which have been used, with good results, in function optimization. However, traditional GAs rapidly push an artificial population toward convergence. That is, all individuals in the population soon become nearly identical. Niching Methods allow genetic algorithms to maintain a population of diverse individuals. GAs that incorporate these methods are capable of locating multiple, optimal solutions within a single population. The purpose of this study is to test existing niching techniques and two methods introduced herein, bearing in mind their eventual application in nuclear reactor related problems, specially the nuclear reactor core reload one, which has multiple solutions. Tests are performed using widely known test functions and their results show that the new methods are quite promising, specially in real world problems like the nuclear reactor core reload. (author)

  13. Desenvolvimento e validação de metodos para a determinação de antimicrobianos em leite e farmacos usando a cromatografia liquida de alta eficiencia e eletroforese capilar

    OpenAIRE

    Monica Cecilia Vargas Mamani

    2007-01-01

    Resumo: Antimicrobianos são largamente empregados na medicina veterinária e resíduos destes podem permanecer nos alimentos de origem animal, acima de valores considerados seguros, quando não são respeitadas as boas práticas veterinárias. O objetivo deste trabalho foi o desenvolvimento e validação de métodos para a determinação de tetraciclinas, sulfonamidas, cloranfenicol e fluoroquinolonas em fármacos usando a eletroforese capilar (CE), assim como método multiresíduos para a determinação de ...

  14. 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.

  15. Comparative analysis between P1 and B1 equations for neutron moderation; Analise comparativa entre os metodos de obtencao e das solucoes das equacoes P1 e B1 para moderacao de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, Aquilino Senra; Silva, Fernando Carvalho da; Cardoso, Carlos Eduardo Santos [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2000-07-01

    In order to calculate the neutron flux in nuclear reactors, B1 or P1 equations are solved by numerical methods for several groups of energy. The neutron fluxes obtained from the solutions of the B1 and P1 equations are similar when they are applied to large nuclear power reactors. However, an important difference between the two fluxes is that the system of P1 equations uses one more approximation than the B1 system and then, its flux is less precise. The present work shows the relations between both equations and analyzes for what conditions the two equations systems are equivalent. Furthermore, this equations are numerically solved in 54 groups of energy for a quadrangular arrange. (author)

  16. Analisis comparativo de una metaheuristica en base a algoritmo genetico vs un metodo de ramificacion y corte para un caso de entrega y recolección con restricciones de ventana de horario

    Directory of Open Access Journals (Sweden)

    Lopez, F.

    2004-07-01

    Full Text Available En la solución de problemas combinatorios, es importante evaluar el costo-beneficio entre la obtención de soluciones de alta calidad en detrimento de los recursos computacionales requeridos. El problema planteado es para el ruteo de un vehículo con entrega y recolección de producto y con restricciones de ventana de horario. En la práctica, dicho problema requiere ser atendido con instancias de gran escala (nodos ≥100. Existe un fuerte porcentaje de ventanas de horario activas (≥90% y con factores de amplitud ≥75%. El problema es NP-hard y por tal motivo la aplicación de un método de solución exacta para resolverlo en la práctica, está limitado por el tiempo requerido para la actividad de ruteo. Se propone un algoritmo genético especializado, el cual ofrece soluciones de buena calidad (% de optimalidad aceptables y en tiempos de ejecución computacional que hacen útil su aplicación en la práctica de la logística. Para comprobar la eficacia de la propuesta algorítmica se desarrolla un diseño experimental el cual hará uso de las soluciones óptimas obtenidas mediante un algoritmo de ramificación y corte sin límite de tiempo. Los resultados son favorables.

  17. Coupling between the differential and perturbation theory methods for calculating sensitivity coefficients in nuclear transmutation problems; Acoplamento entre os metodos diferencial e da teoria da perturbacao para o calculo dos coeficientes de sensibilidade em problemas de transmutacao nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Lubianka Ferrari Russo

    2014-07-01

    The main target of this study is to introduce a new method for calculating the coefficients of sensibility through the union of differential method and generalized perturbation theory, which are the two methods generally used in reactor physics to obtain such variables. These two methods, separated, have some issues turning the sensibility coefficients calculation slower or computationally exhaustive. However, putting them together, it is possible to repair these issues and build a new equation for the coefficient of sensibility. The method introduced in this study was applied in a PWR reactor, where it was performed the sensibility analysis for the production and {sup 239}Pu conversion rate during 120 days (1 cycle) of burnup. The computational code used for both burnup and sensibility analysis, the CINEW, was developed in this study and all the results were compared with codes widely used in reactor physics, such as CINDER and SERPENT. The new mathematical method for calculating the sensibility coefficients and the code CINEW provide good numerical agility and also good efficiency and security, once the new method, when compared with traditional ones, provide satisfactory results, even when the other methods use different mathematical approaches. The burnup analysis, performed using the code CINEW, was compared with the code CINDER, showing an acceptable variation, though CINDER presents some computational issues due to the period it was built. The originality of this study is the application of such method in problems involving temporal dependence and, not least, the elaboration of the first national code for burnup and sensitivity analysis. (author)

  18. Self-adaptive treatment of time dependent nonlinear nonhomogeneous radial heat flow in reactor components with boundary element method; Samoadaptivno obravnanje spemenljivega nelinearnega nehomogenoga radialnega topltnega toka v reaktorskih komponentah z metodo robnih elementov

    Energy Technology Data Exchange (ETDEWEB)

    Sarler, B; Alujevic, A [Univerza B. Kardelja, Institut ' Jozef Stefan' , Ljubljana (Yugoslavia)

    1988-07-01

    The basic principles of self-adaptive algorithm for treatment of transient nonlinear nonhomogeneous radial heat flow, based on direct Boundary Element method formulation, are presented. The indicators of discretization error are developed, together with binary-tree strategy for manipulation with time domain mesh, assuring automatic optimisation of calculation procedure with respect to predetermined error. The developed method is particularly suitable for use in a spectrum of extremely nonlinear cases, occurring in thermal analyses of reactor components.(author)

  19. Boundary element analysis of stress due to thermal shock loading or reactor pressure vessel nozzle; Napetostna analiza pri nestacionarni termicni obremenitvi cevnega prikljucka reaktorske tlacne posode z metodo robnih elementov

    Energy Technology Data Exchange (ETDEWEB)

    Kramberger, J; Potrc, I [Tehniska fakulteta, Maribor (Yugoslavia)

    1989-07-01

    Apart from being exposed to the primary loading of internal pressure and steady temperature field, the reactor pressure vessel is also subject to various thermal transients (thermal shocks). Theoretical and experimental stress analyses show that severe material stresses occur in the nozzle area of the pressure vessel which may lead to defects (cracks). It has been our aim to evaluate these stresses by the use of the Boundary Element method. (author)

  20. Analysis of a calculation method for the determination of the value of safety or control bars; Analisis de un metodo de calculo para la determinacion del valor de barras de seguridad o control

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F; Torres A, C; Filio L, C [ININ, Gcia. de Reactores, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1982-09-15

    Due to the control or safety bars in a nuclear reactor are constituted by strongly absorbent materials, the Diffusion Theory like tool for the calculation of bar values is not directly applicable, should it use the Transport Theory. However the speed and economy of the Diffusion codes for the reactors calculation, those make attractiveness and by this reason its are used in the determination of characteristic parameters and even in the determination of bar values, not without before to make some theoretical developments that allow to make applicable this theory. The application of the Diffusion Theory in strongly absorbent media is based on the use of some effective cross sections distinct from the real ones obtained when imposing the reason that among the flow and it gradient in the external surface of such media (control element in general, bar type or flagstone) be similar to the one obtained using Transport Theory in all the control region (multiplicative and absorbent media) with those real cross sections. The effective cross sections were obtained of the Leopard-NUMICE cell code which has incorporate the respective calculation theory of effective cross sections. Later these constants its were used in the bidimensional diffusion code Exterminator-II, simulating in it, the distribution of safety or control bars. From the cell code its were also obtained the respective constants of the homogeneous fuel cell. The results as soon as those obtained bar values of the diffusion code, its were compared with some experimental results obtained in the R{phi} Swedish reactor of natural uranium and heavy water. In this work an analysis of the bar value of one of them, trying to determine the applicability of the method is made. (Author)

  1. Irradiation alternative method of manganese sulfate solution by a Pu-Be source for efficiency measurements; Metodo alternativo de irradiacao da solucao de sulfato de manganes por uma fonte de Pu-Be para medicoes de eficiencia

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fellipe Souza da; Martins, Marcelo Marques; Pereira, Walsan Wagner, E-mail: fellipess@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    This study intends to create an alternative irradiation system from a Plutonium-Beryllium source for manganese sulphate solution using the Monte Carlo code. Thus seeking to eliminate the issue of institutes that do not have reactors or particle accelerators in its infrastructure, in order to optimize and provide independence for them to carry out efficiency measurements of MnSO{sub 4} solution in their own locality. The Monte Carlo simulations defined the technical features of this new system so that the solution reaches the maximum neutron capture by manganese in solution. (author)

  2. Origin identification for Cantona, Puebla, obsidians by the analysis method of neutron activation (NAA); Identificacion de procedencia para obsidianas de Cantona, Puebla, por el metodo de analisis por activacion neutronica (AAN)

    Energy Technology Data Exchange (ETDEWEB)

    Tellez N, A. L.

    2013-07-01

    There are tests that most of the obsidian worked in the workshops of Cantona, Puebla, is coming from the mineral deposits of Oyameles-Zaragoza, but also has been detected obsidian that macroscopically belongs to other mineral deposits. The present work has as purpose to determine the provenance of an obsidian sample obtained in the Cantona Site to know if there was the presence of obsidian of other mineral deposits. For the study the neutron activation analysis was used to identify the presence of other deposits. An explanation on the treatment to the selected pieces is included, the preparation of the same ones for its irradiation in the nuclear reactor, the counting and statistical study of the results. Finally the results of the selected samples are presented, indicating their origin places, that time comes and the interpretation of the results is given. (Author)

  3. Development of methodology for characterization of cartridge filters from the IEA-R1 using the Monte Carlo method; Desenvolvimento de uma metodologia para caracterizacao do filtro cuno do reator IEA-R1 utilizando o Metodo Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Priscila

    2014-07-01

    The Cuno filter is part of the water processing circuit of the IEA-R1 reactor and, when saturated, it is replaced and becomes a radioactive waste, which must be managed. In this work, the primary characterization of the Cuno filter of the IEA-R1 nuclear reactor at IPEN was carried out using gamma spectrometry associated with the Monte Carlo method. The gamma spectrometry was performed using a hyperpure germanium detector (HPGe). The germanium crystal represents the detection active volume of the HPGe detector, which has a region called dead layer or inactive layer. It has been reported in the literature a difference between the theoretical and experimental values when obtaining the efficiency curve of these detectors. In this study we used the MCNP-4C code to obtain the detector calibration efficiency for the geometry of the Cuno filter, and the influence of the dead layer and the effect of sum in cascade at the HPGe detector were studied. The correction of the dead layer values were made by varying the thickness and the radius of the germanium crystal. The detector has 75.83 cm{sup 3} of active volume of detection, according to information provided by the manufacturer. Nevertheless, the results showed that the actual value of active volume is less than the one specified, where the dead layer represents 16% of the total volume of the crystal. A Cuno filter analysis by gamma spectrometry has enabled identifying energy peaks. Using these peaks, three radionuclides were identified in the filter: {sup 108m}Ag, {sup 110m}Ag and {sup 60}Co. From the calibration efficiency obtained by the Monte Carlo method, the value of activity estimated for these radionuclides is in the order of MBq. (author)

  4. Use of radioanalytical methods for determination of uranium, neptunium, plutonium, americium and curium isotopes in radioactive wastes; Utilizacao de metodos radioanaliticos para a determinacao de isotopos de uranio, plutonio, americio e curio em rejeitos radioativos

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Bianca

    2012-07-01

    Activated charcoal is a common type of radioactive waste that contains high concentrations of fission and activation products. The management of this waste includes its characterization aiming the determination and quantification of the specific radionuclides including those known as Difficult-to-Measure Radionuclides (RDM). The analysis of the RDM's generally involves complex radiochemical analysis for purification and separation of the radionuclides, which are expensive and time-consuming. The objective of this work was to define a methodology for sequential analysis of the isotopes of uranium, neptunium, plutonium, americium and curium present in a type of radioactive waste, evaluating chemical yield, analysis of time spent, amount of secondary waste generated and cost. Three methodologies were compared and validated that employ ion exchange (TI + EC), extraction chromatography (EC) and extraction with polymers (ECP). The waste chosen was the activated charcoal from the purification system of primary circuit water cooling the reactor IEA-R1. The charcoal samples were dissolved by acid digestion followed by purification and separation of isotopes with ion exchange resins, extraction and chromatographic extraction polymers. Isotopes were analyzed on an alpha spectrometer, equipped with surface barrier detectors. The chemical yields were satisfactory for the methods TI + EC and EC. ECP method was comparable with those methods only for uranium. Statistical analysis as well the analysis of time spent, amount of secondary waste generated and cost revealed that EC method is the most effective for identifying and quantifying U, Np, Pu, Am and Cm present in charcoal. (author)

  5. Improved method for lifetime measurements; Methode perfectionnee de mesure de la duree de vie; Usovershenstvovannyj metod izmereniya vremeni zhizni; Metodo perfeccionado para medir la vida media de los estados de excitacion

    Energy Technology Data Exchange (ETDEWEB)

    Weinzierl, P; Bartl, W [Oesterreichische Studiengesellschaft fuer Atomenergie, Seibersdorf (Austria)

    1962-04-15

    (Tl). Pour assurer une grande liberte de disposition geometrique entre les deux scintillateurs, on mesure l'energie gamma apres avoir additionne les amplitudes des impulsions issues de l'un et l'autre detecteur. La precision de fonctionnement du dispositif additionneur est obtenue en faisant d'abord passer les impulsions provenant du cristal organique dans un conditionneur qui ne s'ouvre que s'il y a coincidence entre les deux scintillateurs. Le spectre des impulsions fourni par le dispositif additionneur est transmis a un selecteur a un seul canal pour selection de l'energie gamma. Les rayons gamma (I) sont detectes dans un autre cristal organique. Les impulsions rapides issues des deux scintillateurs organiques sont appliquees a un convertisseur temps-amplitude et a un selecteur multicanaux. Ce selecteur est declenche par coincidence differee entre les discriminateurs d'amplitude des impulsions. (author) [Spanish] Las mediciones de la vida media de los estados excitados de los nucleos se basan en general en la medicion del retardo de un rayo {gamma} (II) respecto de otro rayo {gamma} (I) o de una particula {beta}. Los contadores de centelleo organicos dan el mejor tiempo de resolucion para dicha medida; pero cuando se presentan esquemas complejos de desintegracion, la identificacion de la energia y es importante y la mejor manera de lograrla es mediante detectores de Nal(TI). Para combinar las ventajas de ambos detectores, el rayo {gamma} (II) se hace incidir primeramente sobre un cristal organico (estilbeno) y el quantum de dispersion se detecta en un cristal de Nal(TI). Con el objeto de permitir grandes angulos de admision entre los dos detectores, la medicion de la energia {gamma} se efectua despues de sumar las alturas de los impulsos correspondientes de ambos detectores. Para asegurar el funcionamiento correcto del dispositivo sumador, los impulsos provenientes del cristal organico se hacen pasar primeramente por un circuito de puerta, accionado por cada coincidencia

  6. Different seeds to solve the equations of stochastic point kinetics using the Euler-Maruyama method; Diferentes semillas para solucionar las ecuaciones de la cinetica puntual estocastica empleando el metodo de Euler-Maruyama

    Energy Technology Data Exchange (ETDEWEB)

    Suescun D, D.; Oviedo T, M., E-mail: daniel.suescun@usco.edu.co [Universidad Surcolombiana, Av. Pastrana Borrero - Carrera 1, Neiva, Huila (Colombia)

    2017-09-15

    In this paper, a numerical study of stochastic differential equations that describe the kinetics in a nuclear reactor is presented. These equations, known as the stochastic equations of punctual kinetics they model temporal variations in neutron population density and concentrations of deferred neutron precursors. Because these equations are probabilistic in nature (since random oscillations in the neutrons and population of precursors were considered to be approximately normally distributed, and these equations also possess strong coupling and stiffness properties) the proposed method for the numerical simulations is the Euler-Maruyama scheme that provides very good approximations for calculating the neutron population and concentrations of deferred neutron precursors. The method proposed for this work was computationally tested for different seeds, initial conditions, experimental data and forms of reactivity for a group of precursors and then for six groups of deferred neutron precursors at each time step with 5000 Brownian movements per seed. In a paper reported in the literature, the Euler-Maruyama method was proposed, but there are many doubts about the reported values, in addition to not reporting the seed used, so in this work is expected to rectify the reported values. After taking the average of the different seeds used to generate the pseudo-random numbers the results provided by the Euler-Maruyama scheme will be compared in mean and standard deviation with other methods reported in the literature and results of the deterministic model of the equations of the punctual kinetics. This comparison confirms in particular that the Euler-Maruyama scheme is an efficient method to solve the equations of stochastic point kinetics but different from the values found and reported by another author. The Euler-Maruyama method is simple and easy to implement, provides acceptable results for neutron population density and concentration of deferred neutron precursors and

  7. Correlacion entre metodos de analisis de Zn disponible en cuatro ordenes de suelos de Costa Rica

    Directory of Open Access Journals (Sweden)

    Eloy Molina

    2001-01-01

    Full Text Available Se realizo una comparación entre métodos analisis del Zn disponible en 4 ordenes de sue- Analytilos de Costa Rica (Ultisoles, Vertisoles, Andisoles Inceptisoles, 25 de c/u, utilizando las siguientes soluciones extractoras: Olsen Modificado, Meh- lich 3, Morgan Modificado, DTPA y HC1. Las cantidades de Zn extrafdas dependieron de la natu- raleza qufmica de la solucion extractora. El HCl presento los contenidos mas altos de Zn en los chasuelos, excepto en Vertisoles. Las soluciones que hicontienen el agente quelante EDTA (Olsen Modi- ficado y Mehlich`3, extrajeron niveles interme- Modidios de Zn, en tanto que los metodos que contie- Den el quelato DTPA (Morgan Modificado y DT - PA, obtuvieron los valores mas bajos. Las corre- laciones de Zn extrafble entre los 5 metodos fue- signifirOD significativas en la mayona de los casos, tanto nivel de orden de suelos como en el conjunto de indivilos 100 suelos analizados. Los coeficientes de co- rrelacion mas altos, se presentaron entre Mehlich Morgan Modificado y DTPA. Las correlaciones Modifueron consistentes en los 4 ordenes, 10 que indica que estas soluciones poseen un amplio margen de adaptacion a diferentes tipos de suelo, siendo una caractenstica ventajosa para la selección de un metodo de analisis. El Olsen Modificado fue mas slighteficiente para la extraccion de Zn en suelos de pH ligeramente acido 0 neutro (Vertisoles e Inceptiso- les, que en suelos acidos (Ultisoles y Andisoles. EI HCI extrajo cantidades muy aItas de Zn que Moraparentementestan relacionadas con formas no disponibles para lag plantas. Se concluye que lag soluciones Mehlich 3, Morgan Modificado y DT - PA son semejantes en la forma de extraer Zn dispo- Dible, y podrian seT una altemativa para sustituir el metoda tradicional de Olsen Modificado utilizado en Costa Rica. Sin embargo,la eficiencia de ellas no puede seT establecida sino a traves de log estudios de correlacion contra rendimiento en invernadero y campo.

  8. Proposed method for the hydraulic design of ski-jump energy dissipators in dam spillways considering the occurrence of scour holes downstream of the structure; Metodo propuesto para el diseno hidraulico de trampolines empleados como disipadores de energia en aliviaderos para presas, considerando la ocurrencia del cono de socavacion al pie del mismo

    Energy Technology Data Exchange (ETDEWEB)

    Pardo-Gomez, Rafael [Centro de Investigaciones Hidraulicas (Cuba)

    2008-04-15

    Ski-jump energy dissipators are widely used in hydraulic engineering because of their well-known effectiveness. Nevertheless, some uncertainty exits associated with the dimensions of the scour hole appearing downstream of the structure. This paper presents a new method for solving this problem. This method includes spillway stability checking as part of the design process and also stability checking of any other construction near the energy dissipation zone. [Spanish] Los disipadores de energia tipo trampolin tienen amplia utilizacion en la practica de la ingenieria hidraulica por su probada eficacia; sin embargo, su diseno esta sujeto a cierto grado de incertidumbre en cuanto a la prediccion de las dimensiones del cono de socavacion que habra de producirse aguas abajo de la estructura. En el presente trabajo se muestra un metodo novedoso, mediante el cual el autor soluciona el aspecto antes referido, toda vez que se incluye como parte del proceso de diseno la comprobacion de la estabilidad del propio aliviadero o de cualquier otra obra cercana a la zona de disipacion de energia.

  9. Methods of Particle Detection in Free Neutron Decay; Methode de detection des particules dans une desintegration de neutrons libres; Metod obnaruzheniya chastits pri raspade svobodnogo nejtrona; Metodo para la deteccion de particulas en la desintegracion de neutrones libres

    Energy Technology Data Exchange (ETDEWEB)

    Novey, T B [Argonne National Laboratory, Lemont, IL (United States)

    1960-06-15

    aboutir a un multiplicateur ordinaire a 10 etages et ses chicanes d'entree pour resolution angulaire. 4. Le systeme electronique qui isole dans les detecteurs les impulsions venant des detecteurs et ayant les caracteristiques voulues de frequence, de retard relatif et d'amplitude et permet ainsi d'identifier la des- integration d'un neutron. (author) [Spanish] En el Laboratorio de Argonne se ha llevado a cabo recientemente una serie de estudios experimentales sobre la desintegracion de neutrones polarizados, con el fin de elucidar las modalidades de la interaccion nuclear a bajas energias. Estos estudios han consistido en mediciones de la distribucion angular de electrones y protones con respecto al sentido del spin de los neutrones libres que sufren la desintegracion. Los componentes fundamentales del aparato que el autor describe en la memoria son: 1. Un espejo de hierro y cobalto, de un metro, que selecciona un haz de neutrones altamente polarizados y permite determinar el grado de polarizacion aplicando distintos procedimientos . 2. El detector de electrones, consistente en un mosaico de cristales de antraceno de 10 cm de diametro y 6 mm de espesor, y en un dispositivo de canalizacion luminosa. 3. El detector de protones, formado por un sistema multiplicador de electrones de 14 etapas, la primera de las cuales con una abertura de 15x15 cm, que va disminuyendo en cuatro pasos hasta una estructura multiplicadora usual de 10 etapas, y sus deflectores de entrada para lograr la resolucion angular. 4. El sistema electronico que selecciona los impulsos que, al ser adecuados los intervalos con que se suceden, su defasaje en tiempo y su amplitud, hacen posible la identificacion de una desintegracion neutronica. (author) [Russian] Nedavno Argonnskoj gruppoj byl zavershen ryad ehksperimental'nykh issledovanij po raspadu polyarizovannykh nejtronov s tem, chtoby prolit'svet na strukturu slabogo yadernogo vzaimodejstviya. EHti issledovaniya nosili formu izmereniya uglovykh

  10. 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)

  11. 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.)

  12. 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.)

  13. Estudio numerico y experimental del proceso de soldeo MIG sobre la aleacion 6063--T5 utilizando el metodo de Taguchi

    Science.gov (United States)

    Meseguer Valdenebro, Jose Luis

    improvement on mechanical properties in aluminum metal joint. Los procesos de soldadura por arco electrico representan unas de las tecnicas mas utilizadas en los procesos de fabricacion de componentes mecanicos en la industria moderna. Los procesos de soldeo por arco se han adaptado a las necesidades actuales, haciendose un modo de fabricacion flexible y versatil. Los resultados obtenidos numericamente en el proceso de soldadura son validados experimentalmente. Los principales metodos numericos mas empleados en la actualidad son tres, metodo por diferencias finitas, metodos por elementos finitos y metodo por volumenes finitos. El metodo numerico mas empleado para el modelado de uniones soldadas, es el metodo por elementos finitos, debido a que presenta una buena adaptacion a las condiciones geometricas y de contorno ademas de que existe una diversidad de programas comerciales que utilizan el metodo por elementos finitos como base de calculo. Este trabajo de investigacion presenta un estudio experimental de una union soldada mediante el proceso MIG de la aleacion de aluminio 6063-T5. El metodo numerico se valida experimentalmente aplicando el metodo de los elementos finitos con el programa de calculo ANSYS. Los resultados experimentales obtenidos son: las curvas de enfriamiento, el tiempo critico de enfriamiento t4/3, geometria del cordon, microdurezas obtenidas en la union soldada, zona afectada termicamente y metal base, dilucion del proceso, areas criticas intersecadas entre las curvas de enfriamiento y la curva TTP. Los resultados numericos son: las curvas del ciclo termico, que representan tanto el calentamiento hasta alcanzar la temperatura maxima y un posterior enfriamiento. Se calculan el tiempo critico de enfriamiento t4/3, el rendimiento termico y se representa la geometria del cordon obtenida experimentalmente. La zona afectada termicamente se obtiene diferenciando las zonas que se encuentran a diferentes temperaturas, las areas criticas intersecadas entre las

  14. 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...

  15. 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...

  16. 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%.

  17. 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.

  18. Scienza e conoscenza: sul valore del metodo scientifico

    Directory of Open Access Journals (Sweden)

    Riccardo Luciano Appolloni

    2014-05-01

    Full Text Available L’antico problema di riconoscere una forma di conoscenza oggettivae fondata è ancora vivo; in questo scritto cercheremo di capire se la scienza moderna possa essere una forma di conoscenza tale e, quindi, privilegiata. A tal fine ci serviremo del pensiero di alcuni epistemologi e scienziati. In particolare, nel trattare il problema del valore epistemologico del metodo scientifico, non potremo esimerci dal fare i conti con l’anarchismo metodologico di Paul K. Feyerabend, verso il quale l’esito del presente articolo sarà fondamentalmente critico. A partire dai fecondi spunti di questo filosofo, tenteremo dapprima di analizzare i caratteri distintivi della scienza e del suo metodo rispetto ad altre forme di sapere; quindi, cercheremo di individuare alcuni limiti della conoscenza razionale.

  19. 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

  20. Survey of Pulsed Neutron Source Methods for Multiplying Media; Methodes des Neutrons Pulses Pour l'Etude des Milieux Multiplicateurs; Obzor metodov s ispol'zovaniem istochnikov impul'snykh nejtronov dlya razmnozhayushchej sredy; Estudio Panoramico de los Metodos de Empleo de Fuentes de Neutrons Pulsados en Medios Multiplicadores

    Energy Technology Data Exchange (ETDEWEB)

    Garelis, E. [General Electric Company, Vallecitos Atomic Laboratory, Pleasanton, CA (United States)

    1965-10-15

    -Small-L ). Il discute en outre l'emploi des methodes qui sont fondees sur la reponse a des impulsions pseudoaleatoires et utilisent une correlation entre l'entree et la sortie, pour la determination de la fonction de Green d'un milieu multiplicateur. H montre que les renseignements obtenus par ces methodes sont identiques a ceux fournis par les methodes de la source puisee a repetition, ce qui permet d'appliquer aux premieres les procedes qui ont ete mis au point pour les secondes. (author) [Spanish] En los ultimos afios ha habido dos tendencias principales acerca de la manera mas eficaz de medir la reactividad de parada, empleando fuentes de neutrones pulsados: la primera preconizaba el empleo de metodos tradicionales de medicion con fuentes neutronicas reiteradamente pulsadas, y la segunda propugnaba la aplicacion de metodos basados en una tecnica seudoaleatoria de respuesta a los impulsos, utilizando una correlacion entre los datos de entrada y los de salida. La informacion obtenida con una y otra tecnica es la misma; en teoria, ambos metodos sirven para determinar la funcion de respuesta. Se resefia el desarrollo de las tecnicas de empleo de fuentes de neutrones pulsados aplicadas a sistemas termicos con miras a medir la reactividad, desde los primeros intentos de Sjoestrand hasta el reciente metodo (k{beta}/ Script-Small-L ). Con el metodo usual de empleo de estas fuentes, la propiedad que se procuia determinar es la funcion de Green del conjunto subercritico empleado, es decir, la respuesta del reactor a una fuente de neutrones de funcion delta. El decrecimiento exponencial (e{sup {alpha}t}) de la funcion de Green proporciona una constante de decrecimiento de los neutrones instantaneos que es independiente del espacio. El autor examina los metodos para obtener el valor de la reactividad partiendo de la medida de a, como por ejemplo el metodo de la medicion de la criticidad por neutrones retardados a y el reciente metodo (k{beta}/ Script-Small-L ). Seguidamente examina el

  1. 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...

  2. Comparison Of The Worth Of Critical And Exponential Measurements For Heavy-Water-Moderated Reactors; Valeur Relative des Mesures Critiques et Exponentielles pour l'Etude des Reacteurs Ralentis a l'Eau Lourde; Sravnenie tsennosti kriticheskikh i ehksponentsial'nykh izmerenij dlya reaktorov s tyazhelovodnym zamedlitelem; Valor Relativo de las Mediciones Criticas y Exponenciales para los Reactores Moderados por Agua Pesada

    Energy Technology Data Exchange (ETDEWEB)

    Graves, W. E.; Hennelly, E. J. [Savannah River Laboratory, E.I. Du Pont De Nemours and Co., Aiken, SC (United States)

    1964-02-15

    ultimos diez afios, en el Laboratorio de Savannah River (SRL) se ha venido trabajando con un conjunto critico de agua pesada el PDP, y con un conjunto exponencial, el SE, en forma paralela. Los autores presentan una resefla de a experiencia asi adquirida en dicho laboratorio, exponiendo los resultados y formulando recomendaciones sobre cual de los dos tipos de conjuntos resulta mas adecuado para determinados experimentos. A continuacion se exponen seis tipos de experimentos: 1. Determinacion del laplaciano en reticulados isotropicos uniformes En el SRL se han llevado a cabo extensas comparaciones entre mediciones efectuadas con ayuda de conjuntos criticos de region unica, conjuntos exponenciales, conjuntos criticos de sustitucion y del reactor PCTR (prompt critical test reactor). En el caso de los conjuntos exponenciales, aparentemente solo se tropieza con dificultades en las determinaciones del laplaciano radial. Si se llegase a salvarlas, los experimentos exponenciales podrian rivalizar sin desmedro con los criticos. Los conjuntos criticos de region unica son los que exigen mas material; les sigilen los conjuntos criticos de sustitucion y los exponenciales, cuyas necesidades son, en terminos generales, comparables; en ultimo termino figura el PCTR, cuyas exigencias son minimas. 2. Efectos anisotropicos y de cavitacion Se examinan brevemente loe experimentos realizados en el SRL con conjuntos criticos y se establecen comparaciones entre estos y los conjuntos exponenciales; en otra memoria se trata el mismo tema con mas detalle. ' 3. Evaluacion de los sistemas de control Los experimentos exponenciales se prestan para efectuar mediciones de la eficacia global, siempre que sus resultados se analicen adecuadamente. No obstante, para un estudio cabal de la distribucion general del flujo, de los gradientes de flujo, etc., se requiere un conjunto critico de gran tamano, como el PDP. 4. Coeficientes de temperatura Los experimentos exponenciales constituyen un excelente metodo

  3. 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.

  4. 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...

  5. Flow in potential cascades by means of the finite element method; Flujo en cascadas potenciales mediante el metodo del elemento finito

    Energy Technology Data Exchange (ETDEWEB)

    Sosa Cordero, Rodolfo; Fernandez Valencia, Gonzalo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1987-12-31

    This article presents a mathematical model and its solution by means of the finite element method with approximate Garlekin formulation, for the flow analysis in a circular cascade, in a surface of revolution current of a turbo- machine, that can be axial, mixed or radial. To the revolution surface an agreed transformation is applied to obtain a plane, eliminating in this form one term in the equation succeeding in avoiding an iterative solution. Likewise, the finite element method allows to solve the equation in partial derivatives of the elliptical type in its quasi-harmonic form. Additionally, the method followed to introduce the contour conditions is presented; specially, the Kutta-Joukowsky conditions and the one of periodicity, which distinguishes this problem from the classical problems of ideal flows evaluated in the contour. [Espanol] En este articulo se presenta un modelo matematico y su solucion mediante el empleo del metodo del elemento finito con formulacion aproximada de Galerkin, para el analisis del flujo en una cascada circular, en una superficie de corriente de revolucion de una turbomaquina, que puede ser axial, mixta o radial. A la superficie de revolucion se le aplica una transformacion conforme para obtener un plano, eliminando de esta forma un termino en la ecuacion logrando evitar la solucion iterativa. Asimismo, el metodo del elemento finito permite resolver la ecuacion en derivadas parciales del tipo eliptico en su forma cuasiarmonica. Se presenta, ademas, el metodo seguido para introducir las condiciones de contorno; en especial, las condiciones de Kutta-Joukowsky y la de periodicidad, que distinguen a este problema de los problemas clasicos de flujos ideales valuados en el contorno.

  6. Flow in potential cascades by means of the finite element method; Flujo en cascadas potenciales mediante el metodo del elemento finito

    Energy Technology Data Exchange (ETDEWEB)

    Sosa Cordero, Rodolfo; Fernandez Valencia, Gonzalo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1986-12-31

    This article presents a mathematical model and its solution by means of the finite element method with approximate Garlekin formulation, for the flow analysis in a circular cascade, in a surface of revolution current of a turbo- machine, that can be axial, mixed or radial. To the revolution surface an agreed transformation is applied to obtain a plane, eliminating in this form one term in the equation succeeding in avoiding an iterative solution. Likewise, the finite element method allows to solve the equation in partial derivatives of the elliptical type in its quasi-harmonic form. Additionally, the method followed to introduce the contour conditions is presented; specially, the Kutta-Joukowsky conditions and the one of periodicity, which distinguishes this problem from the classical problems of ideal flows evaluated in the contour. [Espanol] En este articulo se presenta un modelo matematico y su solucion mediante el empleo del metodo del elemento finito con formulacion aproximada de Galerkin, para el analisis del flujo en una cascada circular, en una superficie de corriente de revolucion de una turbomaquina, que puede ser axial, mixta o radial. A la superficie de revolucion se le aplica una transformacion conforme para obtener un plano, eliminando de esta forma un termino en la ecuacion logrando evitar la solucion iterativa. Asimismo, el metodo del elemento finito permite resolver la ecuacion en derivadas parciales del tipo eliptico en su forma cuasiarmonica. Se presenta, ademas, el metodo seguido para introducir las condiciones de contorno; en especial, las condiciones de Kutta-Joukowsky y la de periodicidad, que distinguen a este problema de los problemas clasicos de flujos ideales valuados en el contorno.

  7. Evaluation and standardization of neutron activation analysis according to the K{sub 0} method in the RP-10 reactor; Evaluacion y estandarizacion del analisis por activacion neutronica segun el metodo del K{sub 0} en el reactor nuclear RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Montoya R, E

    1995-06-01

    It has been characterized and standardized an irradiation of the RP-10 Research Nuclear Reactor for use of the K{sub 0} method of neutron activation analysis using the Hoegdahl convention; also it has been evaluate the behaviour of such method in regard to the accuracy and precision of the results obtained in the quantitative multi elemental analysis of several certified materials of reference. In order to prove that the analytical method is totally under statistical control, it has been used the Heydorn method. It has been verified that the method is exact, precise and reliable to determine the aluminium, antimuonium, arsenic, bromine, calcium, chloride, copper, magnesium, manganese, sodium, titanium, vanadium, zinc and other elements. Also, they are discussed, in regard to the use of K{sub 0} constants, the different formalisms employed to calculate the integral of the reaction rate by nucleus in the activation. (author). 58 refs., 18 tabs., 6 figs.

  8. 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.

  9. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  10. Some Possibilities of the Eddy-Current Method for Multi-Parameter Testing of Structural Components; Quelques Possibilites Offertes par la Methode des Courants de Foucault pour le Controle de Nombreux Parametres des Elements de Construction; Nekotorye vozmozhnosti metoda vikhrevykh tokov dlya mnogoparametrovogo kontrolya ehlementov konstruktsij; Algunas Posibilidades que Brinda el Metodo de las Corrientes de Foucault para Controlar Numerosos Parametros de los Elementos de Construccion

    Energy Technology Data Exchange (ETDEWEB)

    Vjahorev, V. G.; Gerasimov, V. G.; Deniskin, V. P.; Trahtenberg, L. I.; Shkarlet, Ju. M. [Gosudarstvennyj Komitet po Ispol' zovaniju Atomnoj Jenergii SSSR, Moskva, SSSR (Russian Federation)

    1965-09-15

    donnees calculees et experimentales, les auteurs indiquent diverses possibilites de realisation de detecteurs a courants de Foucault de frequence unique permettant le controle avec modification simultanee de plusieurs parametres. (author) [Spanish] Los autores demuestran que los problemas del control no destructivo de diversos parametros se plantean en la tecnologia nuclear segun ciertas leyes y que el metodo de las corrientes de Foucault se adapta particularmente a la solucion de varios de esos problemas. Los autores justifican la utilizacion de modelos electricos para resolver problemas de control con ayuda de una larga bobina hueca. Presentan formulas y exponen un metodo para el calculo de detectores a base de corrientes parasitas. Describen un dispositivo que permite verificar el espesor de las paredes de tubos; en este dispositivo, el desplazamiento del tubo a controlar no ejerce influencia alguna sobre los resultados de las mediciones, gracias al empleo de un monitor accionado por una senal dependiente de la fase de tension del detector. En calidad de detector se ha empleado un dispositivo destinado al control de los tubos de un autogenerador despues de haber incluido en su circuito una bobina de ensayo. Partiendo de datos calculados y experimentales, los autores senalan diversas posibilidades de realizacion de detectores a base de corrientes de Foucault de frecuencia unica, que permiten efectuar el control con modificacion simultanea de varios parametros. (author) [Russian] Pokazana zakonomernost' voznikno- venija zadach o nerazrushajushhem mnogoparametrovom kontrole v jadernoj tehnologii i celeso- obraznost' reshenija nekotoryh iz nih metodom vihrevyh tokov. Obosnovano primenenie jelektricheskih modelej dlja reshenija zadach kontrolja s ispol'zovaniem dlinnoj prohodnoj katushki. Privodjatsja raschetnye formuly i izlagaetsja metodika rascheta nakladnyh toko- vihrevyh datchikov. Opisana shema pribora dlja kontrolja tolshhiny stenki trub, v kotorom vlijanie pereme

  11. Flow measurement in a 170-MW hydraulic turbine using the Gibson method; Medicion del flujo de una turbina hidraulica de 170 MW utilizando el metodo Gibson

    Energy Technology Data Exchange (ETDEWEB)

    Urquiza, Gustavo [Universidad Autonoma del Estado de Morelos (Mexico); Adamkowski, Adam [The Szewalski Institute of Fluid-Flow Machinery (Poland); Kubiak, Janusz; Sierra, Fernando [Universidad Autonoma del Estado de Morelos (Mexico); Janicki, Waldemar [The Szewalski Institute of Fluid-Flow Machinery (Poland); Fernandez, J. Manuel [Comision Federal de Electricidad (Mexico)

    2007-07-15

    This paper describes the methodology applied for measuring water flow through a 170-MW hydraulic turbine. The flow rate was measured using the pressure-time method, also known as the Gibson method. This method uses the well-known water hammer phenomenon in pipelines; in turbine penstocks, for instance. The version of this method used here is based on measuring, during total stop of the water stream, the time-history of pressure change in one section of the turbine penstock and relate it to the pressure in the upper reservoir to which the penstock is connected. The volumetric flow rate is determined from the relevant integration of the measured temporary pressure rise. Flow measurement was possible this way because the influence of the penstock inlet was negligible as far as an error of the measurement is concerned. The length of the penstock was 300 m. Previous experience and a standard IEC-41-1991 were the criteria adopted and applied. A fast and efficient acquisition system, including a 16 bit card, was used. The flow rate was calculated using a computer program developed and tested on several cases. The results obtained with the Gibson method were used for calibration of the on-line flow measuring system based on the Winter-Kennedy method as one of the index methods. This method is very often used for continuous monitoring of the flow rate through hydraulic turbines, when the calibration has been done on site by using the results of measurements obtained by the absolute method. Having measured the flow rate and output power, the efficiency was calculated for any operating conditions. A curve showing the best operating conditions based on the highest efficiency is presented and discussed. The details of the instrumentation, its installation, and the results obtained are discussed in the paper. [Spanish] Este articulo describe la metodologia aplicada para la medicion del flujo en una turbina hidraulica de 170 MW. El flujo se midio utilizando el metodo de presion

  12. Reactor Physics Development for Advanced Gas-Cooled Reactors; Recherches en Physique des Reacteurs, pour des Reacteurs Perfectionnes Refroidis par un Gaz; Razrabotka metodov v oblasti reaktornoj fiziki dlya usovershenstvovannogo reaktora s gazovym okhlazhdeniem; Progresos de la Fisica de los Reactores de Tipo Avanzado Refrigerados por Gas

    Energy Technology Data Exchange (ETDEWEB)

    Moore, J. [United Kingdom Atomic Energy Authority (United Kingdom)

    1964-04-15

    para la produccion de energia nucleoelectrica con fines civiles. La principal finalidad de la memoria es describir el considerable trabajo que entrano el desarrollo de metodos teoricos adecuados para calcular: a) la distribucion del flujo y el balance de la reactividad en un cuerpo complejo, b) la distribucion de la potencia en geometrias complejas del combustible, y c) el efecto de la irradiacion sobre los ciclos del combustible y la distribucion de la potencia. A modo de introduccion se menciona la informacion experimental y los metodos teoricos que constituye el resultado de los trabajos con sistemas uranio-magnox, y los datos experimentales comunicados por el British Industries Collaborative Experimental Programma (BICEP), en los que se baso el desarrollo de los metodos teoricos que se han aplicado a los reactores AGR. Con el fin de determinar loe parametros del reticulado del AGR y comprobar los metodos teoricos establecidos para cuerpos de reactor heterogeneos, se ha empleado el conjunto critico APEX y el reactor HERO de energia nula, tanto con reticulados normales como con combinaciones de perturbadores tales como barras de control. Los metodos teoricos desarrollados y empleados hasta ahora se conocen por el nombre de 'hetrecontrol' y 'FTD2'. Se prepararon experimentos para comprobar algunos detalles de las caracteristicas de estos metodos y se han analizado mediciones efectuadas en las instalaciones APEX y HERO con varios cuerpos de 'reactor' de diversos tamanos con el fin de determinar series coherentes de constantes reticulares que concuerden con los resultados experimentales. Seguidamente, a estas constantes puramente empiricas se aplicaron los metodos 'hetrecontrol' y 'FTD2' para planear la puesta en marcha y elegir el esquema de carga del reactor AGR de Windscale. La memoria menciona las tecnicas experimentales comprobadas y las que se han desarrollado para resolver los problemas particulares que se presentaron. Reviste particular interes el examen de los

  13. Torque calculation in the induction motor with the finite element method; Calculo del par en el motor de induccion con el metodo del elemento finito

    Energy Technology Data Exchange (ETDEWEB)

    Castillo Diaz, Ramon

    2002-06-15

    In this work the method of the finite element is applied to the bi-dimensional analysis of the induction motor in operation in steady state, excited by sine sources of laminar currents and sine sources of voltage. The analysis is focused mainly in the calculation of the electromagnetic torque. The topics of electromagnetic theory are covered and in an idealized model of the induction motor, analytically and numerically with the method of the finite element, in the variant method of Galerkin, the vectorial potential and the torque are calculated. The results obtained with the analytical and numerical methods are compared. Three formulations are developed to calculate the torque with the method of the finite element, using triangular elements of first order, based in the equation of force of Lorentz, the Maxwell tensor and the principle of the virtual work. Finally, a motor of induction of real characteristics is simulated, assuming it is connected to a three-phase voltage source. In this motor it is analyzed the convergence and the evolution in the results obtained of the torque with different discretions, and the torque-velocity performance curve is calculated. [Spanish] En este trabajo se aplica el metodo del elemento finito al analisis bidimensional del motor de induccion en operacion en estado estable, excitado por fuentes de corriente laminar senoidales y fuentes de voltaje senoidales. El analisis se enfoca principalmente en el calculo del par electromagnetico. Se tratan los topicos de teoria electromagnetica involucrados y en un modelo idealizado del motor de induccion, se calculan analitica y numericamente con el metodo del elemento finito, en la variante metodo de Galerkin, el potencial vectorial y el par. Se comparan resultados obtenidos con los metodos analiticos y numericos. Se desarrollan tres formulaciones para calcular el par con el metodo del elemento finito, utilizando elementos triangulares de primer orden, basadas en la ecuacion de fuerza de

  14. 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

  15. A Method of Identification and Inspection for Inventory Control of Irradiated Fuel Elements; Methode d'Identification et d'Inspection Permettant de Proceder a l'Inventaire des Elements Combustibles Irradies; Metod identifikatsii i proverki pri inventarnom kontrole obluchennykh toplivnykh ehlementov; Metodos de Identificacion e Inspeccion para el Control de las Existencias de Elementos Combustibles Irradiados

    Energy Technology Data Exchange (ETDEWEB)

    Kinderman, E. M.; Mills, J. S. [Stanford Research Institute, Menlo Park, CA (United States)

    1966-02-15

    punto de descarga del reactor y el de almacenamiento del combustible. Pero esos dos lugares presentan dificultades materiales de caracter peculiar para los inspectores o los encargados de controlar el inventario. La distancia entre el objeto y el observador, la escasa iluminacion y l a distorsion optica debida a los medios de blindaje (agua, cristal) son factores que contribuyen a dificultar las operaciones inherentes al inventario de materiales. Para resolver estas dificultades y poder identificar positivamente el combustible descargado de un reactor, los autores han ideado un dispositivo de inspeccion optica compuesto de una estacion fija, un periscopio, un telescopio y una camara fotografica. Con este dispositivo se han hecho pruebas en una pileta de irradiacion por cobalto en las que se han examinado varias muestras marcadas diversamente y tratadas en un circuito de agua caliente. Para estas pruebas se emplearon un periscopio lleno de agua, de 3,35 m de longitud, un telescopio catadioptrico colocado a 4,9 m del periscopio, y una camara de 35 mm para registrar las observaciones. El telescopio, la camara y los soportes son portatiles, pues pesan en total menos de 10 kg y su dimension maxima es de 75 cm. Las observaciones fotograficas de diagramas de resolucion tomadas como pruebas en aire, a 6 m del objetivo del telescopio, pusieron de manifiesto que ese sistema puede resolver senales de comprobacion de 22 {mu}m de ancho, lo que corresponde a una resolucion de 0,8 segundos de arco. Ensayos llevados a cabo con el dispositivo sumergido en agua mostraron que en este caso el poder de resolucion es superior a 50 {mu}m a 6 m, lo que corresponde a una resolucion de 1 segundo de arco, Es evidente que el poder de resolucion nunca puede ser superior al del sistema en aire. El sistema descrito empleado en un inventario real de combustible, reproducira adecuadamente cualquier senal de comprobacion. (author) [Russian] Mesta razgruzki reaktora i hranenija obluchennogo topliva

  16. Experience with a New Colour-Scintillographic Method for Diagnosing Liver Tumours and Inflammatory Liver Disorders; Experience d'une Nouvelle Methode d'Enregistrement Scintigraphique en Couleurs, pour le Diagnostic des Tumeurs Hepatiques et des Affections Inflammatoires du Foie; Opyt ispol'zovaniya novogo metoda tsvetnoj'' stsintigrafii dlya diagnostiki pechenochnykh opukholej i vospalitel'nykh protsessov pecheni; Ensayo de un Nuevo Metodo de Registro Centelleografico Policromo, para el Diagnostico de Tumores Hepaticos y de Inflamaciones del Higado

    Energy Technology Data Exchange (ETDEWEB)

    Sparchez, T.; Gheorghescu, B.; Steclaci, A.; Merculiev, E.; Popovici, M. [Clinique Medicale, Bucarest (Romania)

    1964-10-15

    que le noir represente l'activite maximum, situee au centre du foie ou le parenchyme est plus epais. Les couleurs intermediaires correspondent aux differentes zones d'isoradioactivite. Les auteurs ont etudie, a l'aide de cette methode, 150 cas de tumeurs hepatiques malignes (primitives et secondaires) et de tumeurs benignes, 80 cas d'hepatite chronique et de cirrhose et 50 cas normaux. Dans la majorite descas, ils ont utilise l'or-198 colloiedal (Amersham-Angleterre) injecte par voie intraveineuse. Ils ont effectue parallelement, dans un grand nombre de cas, des mecanogrammes, photoscintigrammes en blanc et noir et scintigrammes en couleurs. Le diagnostic a ete verifie par ponction, laparoscopie, laparophoto ou cinematographie, ponction biopsique, intervention chirurgicale ou necropsie. La methode scintigraphique en couleurs permet de mieux distinguer les variations d'intensite de la radioactivite, c'est-a-dire la desorganisation ou la substitution du parenchyme hepatique par processus tumoraux. (author) [Spanish] Para obtener imagenes que reflejasen mejor los detalles de la estructura del parenquima hepatico modificado por la enfermedad, los autores adaptaron al aparato Scanner-Tracerlab un dispositivo gracias al cual se pueden obtener diagramas hepaticos en siete colores. Cada color, elegido arbitrariamente, corresponde a un numero dado de impulsos y representa zonas de isorradiactividad, es decir, de tejido hepatico de volumen relativamente igual. En el centelleograma policromo, el blanco representa la radiactividad de fondo, mientras que el negro representa la actividad maxima, situada en el centro del higado, donde el parenquima es mas espeso. Los colores intermedios corresponden a las diferentes zonas de isorradiactividad. Recurriendo a este metodo, los autores estudiaron 150 casos de tumores hepaticos malignos (primitivos y secundarios) y de tumores benignos, 80 casos de hepatitis cronica y de cirrosis, y 50 casos normales. En la mayoria de los casos utilizaron

  17. 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.

  18. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  19. 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.

  20. A modification of the method for determining current efficiency of aluminium electrolytic cells; Modification de la methode permettant de determiner le rendement des cuves dans la production d'aluminium par electrolyse; Izmenenie metoda opredeleniya ehffektivnosti toka v alyuminievykh ehlektroliticheskikh bakakh; Modificacion del metodo para determinar el rendimiento de las celdas utilizadas en la produccion de aluminio por electrolisis

    Energy Technology Data Exchange (ETDEWEB)

    Pradzynski, A [Institute of Basic Technical Problems, Polish Academy of Sciences. Warsaw (Poland); Orman, Z [Institute of Nonferrous Metals, Gliwice (Poland)

    1962-01-15

    de faciliter l'application de cette methode dans les usines d'aluminium et d'eviter toutes les restrictions et tous les dangers qu'entraine la manipulation, en dehors des laboratoires speciaux pour l'etude des radioisotopes, de sources radioactives non scellees. De l'or inactif a ete introduit dans' l'alliage type et dans le bain de la cuve electrolytique. La concentration d'or dans les echantillons d'alliage type et dans les echantillons preleves dans la cuve a ete mesuree apres irradiation de ces echantillons dans un reacteur nucleaire. (author) [Spanish] El procedimiento para determinar el rendimiento de las celdas para la produccion de aluminio por electrolisis fue descrito inicialmente por Rempel y col. y fue perfeccionado por Bozoky y col. que emplearon el radioisotopo {sup 198}Au. Este procedimiento consiste en preparar aleaciones tipo de aluminio con {sup 198}Au y medir la elevada actividad especifica de muestras de la aleacion con un tubo Geiger-Mueller, introduciendo plomo como absorbente entre el tubo y la muestra. Los autores midieron la actividad especifica de la aleacion tipo despues de diluirla con una cantidad conocida de aluminio puro. De esta manera, las muestras de aleacion tipo diluida y las muestras tomadas en la celda electrolitica tienen una actividad especifica del mismo orden de magnitud, que puede ser medida sin necesidad de absorbente. Los autores han recurrido al analisis por radiactivacion con objeto de facilitar la aplicacion de este procedimiento en las fabricas de aluminio y evitar las restricciones y los peligros que supone la utilizacion de fuentes de radiacion no encerradas fuera .de los laboratorios de radioisotopos. El procedimiento consiste en introducir oro inactivo en la aleacion tipo y en la masa fundida de la celda electrolitica. Se extraen muestras de la aleacion y del electrolito y, despues de irradiarlas en un reactor nuclear, se determina la concentracion de oro. (author) [Russian] Metod opredeleniya ehffektivnosti toka v

  1. Calibration of SPND/Rhodium device for mapping the neutron fluence in the IEA-R1 reactor by means of the activation foil method; Calibracao de um dispositivo de mapeamento de fluxo de neutrons - SNPD/Rodio no reator IEA-R1, por meio do metodo de ativacao de folhas

    Energy Technology Data Exchange (ETDEWEB)

    Ricci Filho, Walter; Dias, Mauro S.; Tondin, Julio B.M.; Koskinas, Marina F. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    The IEA-R1 reactor has undergone a modernization to increase its operating power to 5 MW, in order to allow a more efficient production of radioisotopes. The objective of this work is to provide the reactor with flux monitoring device using a rhodium Self-Powered Neutron Detector (SPND). The work presents the results obtained with Rhodium-SPND in several irradiation positions inside the reactor core. A calibration procedure has been performed by means of {sup 197} Au activation foils, with and without cadmium cover, in order do measure the thermal and epithermal neutron fluxes. (author)

  2. 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.

  3. 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)

  4. 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.

  5. Application of the nodal method RTN-0 for the solution of the neutron diffusion equation dependent of time in hexagonal-Z geometry; Aplicacion del metodo nodal RTN-0 para la solucion de la ecuacion de difusion de neutrones dependiente del tiempo en geometria hexagonal-Z

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J.; Alonso V, G. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: jaime.esquivel@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)

    2015-09-15

    The solution of the neutron diffusion equation either for reactors in steady state or time dependent, is obtained through approximations generated by implementing of nodal methods such as RTN-0 (Raviart-Thomas-Nedelec of zero index), which is used in this study. Since the nodal methods are applied in quadrangular geometries, in this paper a technique in which the hexagonal geometry through the transfinite interpolation of Gordon-Hall becomes the appropriate geometry to make use of the nodal method RTN-0 is presented. As a result, a computer program was developed, whereby is possible to obtain among other results the neutron multiplication effective factor (k{sub eff}), and the distribution of radial and/or axial power. To verify the operation of the code, was applied to three benchmark problems: in the first two reactors VVER and FBR, results k{sub eff} and power distribution are obtained, considering the steady state case of reactor; while the third problem a type VVER is analyzed, in its case dependent of time, which qualitative results are presented on the behavior of the reactor power. (Author)

  6. Factors application of MW-mile method and participation in the allocation of charges by the drills use in electricity markets; Aplicacion de los factores de participacion y del metodo de MW-milla en la asignacion de cargos por uso de redes de transmision en mercados de electricidad

    Energy Technology Data Exchange (ETDEWEB)

    Alba-Gomez, L; Tovar-Hernandez, J. H; Gutierrez-Alcaraz, G [Instituto Tecnologico de Morelia, Michoacan (Mexico)]. E-mail: horaciotovar@mexico.com; ggutier@itmorelia.edu.mx

    2007-04-15

    Use of network allocation costs by shift factors and MW-Mile method is reported in this paper. Conventional shift factors are computed based on DC power flow. DC power flow requires to selecting a slack bus in order to avoid matrix singularity. Therefore, shift factors are slack bus dependent. In order to evade slack bus dependency, two approaches are considered. [Spanish] Este trabajo presenta la asignacion de costos por uso de red mediante la aplicacion de Factores de Participacion (FP) y el metodo de MW-Milla. Los FP clasicos son calculados a partir del modelo lineal de flujos de potencia para lo que se requiere de establecer un nodo de referencia a fin de eliminar la singularidad de la matriz de coeficientes. Por lo tanto, los FP son dependientes de la asignacion del nodo de referencia. Dos metodos alternativos para evitar la dependencia del nodo compensador en la obtencion de los factores de participacion son presentados.

  7. 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.

  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. 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

  10. LE «SOTTIGLIEZZE DI CERTA DIDATTICA SUPERLATIVA» DELLA GRAMMATICA ELEMENTARE: STORIA (ATTESTATA DEL METODO RAFFORZISTA (1814-1914

    Directory of Open Access Journals (Sweden)

    Michela Dota

    2018-03-01

    Full Text Available Il contributo ripercorre la storia e le peculiarità del metodo rafforzista, metodo glottodidattico per l’insegnamento della lettura e della scrittura. Il metodo, nato in Italia nel primo Ottocento, nell’epoca postunitaria era praticato nelle scuole elementari soprattutto dell’Italia meridionale, nonché nelle scuole reggimentali e in alcuni istituti per sordomuti. Le sue fondamenta, aberranti rispetto alla norma ortografica e ortoepica tradizionale, lo resero obiettivo di un tenace ostracismo da parte del Ministero dell’Istruzione pubblica, sostenuto per questa occasione da due tra i più eminenti glottologi dell’epoca: Graziadio Isaia Ascoli e Francesco Lorenzo Pullè. Il metodo finì per estinguersi nel secondo decennio del Novecento.   The history of the “metodo rafforzista” for teaching Italian (1814-1914 This article retraces the history and peculiarities of the “metodo rafforzista”, a language teaching method developed in Italy during the first part of 1800s. In the post-Unitarian period, it was used in elementary schools, especially in Southern Italy, and also in military schools and in some institutes for the Deaf and Dumb. Orthographic and orthoepic models proposed by the “metodo rafforzista” diverged from rules of traditional Italian grammar. The method, supported by Graziadio Isaia Ascoli and Francesco Lorenzo Pullè, two of the most distinguished Italian linguists at that time, was subsequently rejected by the Ministry of Public Education, and it vanished during the second half of 1900s.

  11. 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...

  12. Alternative method of portable irradiation of manganese sulphate solution by an plutonium-beryllium source for manganese sulphate bath efficiency measurements; Metodo alternativo de irradiacao portatil da solucao de sulfato de manganes por uma fonte de plutonio-berilio para medicoes de eficiencia do banho de sulfato de manganes

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fellipe Souza da; Martins, Marcelo Marques; Pereira, Walsan Wagner, E-mail: fellipess@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2016-07-01

    This study intends to create an alternative irradiation system from a Plutonium-Beryllium source for manganese sulphate solution using the Monte Carlo code. Thus seeking to eliminate the issue of institutes that do not have reactors or particle accelerators in its infrastructure, in order to optimize and provide independence for them to carry out efficiency measurements of MnSO{sub 4} solution in their own locality. The Monte Carlo simulations defined the technical features of this new system so that the solution reaches the maximum neutron capture by manganese in solution. (author)

  13. 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%.

  14. The method of the maximum entropy for the reconstruction of the distribution bolt the bolt of the neutrons flow in a fuel element; O metodo da maxima entropia para a reconstrucao da distribuicao pino a pino do fluxo de neutrons em um elemento combustivel

    Energy Technology Data Exchange (ETDEWEB)

    Ancalla, Lourdes Pilar Zaragoza

    2005-04-15

    The reconstruction of the distribution of density of potency pin upright in a heterogeneous combustible element, of the nucleus of a nuclear reactor, it is a subject that has been studied inside by a long time in Physics of Reactors area. Several methods exist to do this reconstruction, one of them is Maximum Entropy's Method, that besides being an optimization method that finds the best solution of all the possible solutions, it is a method also improved that uses multipliers of Lagrange to obtain the distribution of the flows in the faces of the combustible element. This distribution of the flows in the faces is used then as a contour condition in the calculations of a detailed distribution of flow inside the combustible element. In this work, in first place it was made the homogenization of the heterogeneous element. Soon after the factor of the multiplication executes and the medium values of the flow and of the liquid current they are computed, with the program NEM2D. These values medium nodal are, then, used upright in the reconstruction of the distribution pin of the flow inside the combustible element. The obtained results were acceptable, when compared with those obtained using fine mesh. (author)

  15. Use of small reactors as an alternative to supply electricity to Baja California Sur; Uso de reactores pequenos como alternativa de suministro de electricidad para Baja California Sur

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, G.; Portes, E.; Ramirez, J. R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ortega, G., E-mail: gustavo.alonso@inin.gob.mx [Comision Federal de Electricidad, Rio Rodano No. 14, 06500 Ciudad de Mexico (Mexico)

    2016-09-15

    The state of Baja California Sur (Mexico) does not form part of the national interconnected electrical system of the country, reason why is local its electrical power supply; one of the alternatives to cover future demands is the use of gas-based combined cycles, which presents the additional problem of including a high price for gas transportation in its costs. In order to reduce total costs, including investment, fuels and operation and maintenance in the operation of the Baja California Sur state electricity system in the coming years, mainly due to the estimated natural gas cost order of $11.50 dollars per million BTU, a proposal is presented to reduce the costs of the electrical system by replacing the necessary combined cycles with the new Small Modular Reactor type nuclear reactors, this alternative is economically competitive. (Author)

  16. 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)

  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. 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)

  19. Preliminary Validation and Verification Plan for CAREM Reactor Protection System; Modelo de Plan Preliminar de Validacion y Verificacion para el Sistema de Proteccion del Reactor CAREM

    Energy Technology Data Exchange (ETDEWEB)

    Fittipaldi, Ana; Felix, Maciel [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

    2000-07-01

    The purpose of this paper, is to present a preliminary validation and verification plan for a particular architecture proposed for the CAREM reactor protection system with software modules (computer based system).These software modules can be either own design systems or systems based in commercial modules such as programmable logic controllers (PLC) redundant of last generation.During this study, it was seen that this plan can also be used as a validation and verification plan of commercial products (COTS, commercial off the shelf) and/or smart transmitters.The software life cycle proposed and its features are presented, and also the advantages of the preliminary validation and verification plan.

  20. Design of an anaerobic hybrid reactor for industrial wastewater treatment; Diseno de reactores hibridos anaerobios para el tratamiento de aguas residuales industriales

    Energy Technology Data Exchange (ETDEWEB)

    Soroa del Campo, S.; Lopetegui Garnika, J.; Almandoz Peraita, A.; Garcia de las Heras, J. L.

    2005-07-01

    The application of the European legislation has promoted different strategies aimed at minimizing the biological sludge production during wastewater treatment. Anaerobic biological treatment is the clearest choice from a technical and economical point of view regarding industrial wastewater. In this context, a semi-industrial anaerobic hybrid reactor has been developed as an alternative technology to other anaerobic systems well-established in the market for the treatment of slaughterhouse wastewater. The The results have demonstrated that it is an effective, robust and easy to operate system. The sludge production has been reduced below 0.12 kg VS/kg COD removed, for COD removal efficiencies above 95%. (Author) 12 refs.

  1. Deteccion De Los Diferentes Grados De Motricidad A Traves Del Metodo De Mabc Para Encontrar Las Estrategias Metodologicas De Enseñanza A Los Niños Con Sindrome De Down, En La Escuela Especial “Un Nuevo Amanecer”, De La Ciudad De Babahoyo

    OpenAIRE

    Galarza Acosta, Fresia

    2015-01-01

    Detectar dificultades en la aplicación de estrategias metodológicas de enseñanza mediante la evaluación del desarrollo de habilidades motrices y destrezas cognitivas que predominan en los niños con síndrome de Down que acuden a la escuela especial Un Nuevo Amanecer mediante la comparación con los parámetros del test de Mabc para conocer cuáles son las estrategias más adecuadas de enseñanza. Se hacen esfuerzos por incluir en la sociedad a las personas con síndrome de Down, especial atenció...

  2. Development of neutronic models with KANEXT for a reactor of traveling wave; Desarrollo de modelos neutronicos con KANEXT para un reactor de onda viajera

    Energy Technology Data Exchange (ETDEWEB)

    Lopez S, R. C.; Francois L, J. L. [UNAM, Facultad de Ingeniera, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Becker, M., E-mail: rcarlosls@yahoo.com.mx [Institut fur Neutronenphysik und Reaktortechnik (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2012-10-15

    Problems with the computation time in a based code on the Monte Carlo method, they conduct to explore the option of the deterministic codes to be able to develop a model of robust reactor of traveling wave, and where short term results are obtained to carry out experimentation work to the moment to study an assemblies exchange scheme as method of fuel administration. KANEXT is a versatile and reliable code, developed in Germany that has satisfied our development necessities until the moment. In this article is described the KANEXT operation, and like it was implemented to develop a preliminary model of reactor core of traveling wave with operation way of stationary wave. The results obtained until the moment, as for the neutrons multiplication factor and the isotopic balance, are encouraging and they lead to refine the model removing completely the axial covering. The adoption of deterministic codes has allowed carrying out tests of complete core in a conventional computer in question of only hours; this will be valuable in the next stage of the research, where developing a re situate scheme of fuel assemblies will involve a great quantity of sprints. (Author)

  3. AZTLAN platform: Mexican platform for analysis and design of nuclear reactors; AZTLAN platform: plataforma mexicana para el analisis y diseno de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Gomez T, A. M.; Puente E, F. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edif. 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Francois L, J. L.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: armando.gomez@inin.gob.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2014-10-15

    The Aztlan platform Project is a national initiative led by the Instituto Nacional de Investigaciones Nucleares (ININ) which brings together the main public houses of higher studies in Mexico, such as: Instituto Politecnico Nacional, Universidad Nacional Autonoma de Mexico and Universidad Autonoma Metropolitana in an effort to take a significant step toward the calculation autonomy and analysis that seeks to place Mexico in the medium term in a competitive international level on software issues for analysis of nuclear reactors. This project aims to modernize, improve and integrate the neutron, thermal-hydraulic and thermo-mechanical codes, developed in Mexican institutions, within an integrated platform, developed and maintained by Mexican experts to benefit from the same institutions. This project is financed by the mixed fund SENER-CONACYT of Energy Sustain ability, and aims to strengthen substantially to research institutions, such as educational institutions contributing to the formation of highly qualified human resources in the area of analysis and design of nuclear reactors. As innovative part the project includes the creation of a user group, made up of members of the project institutions as well as the Comision Nacional de Seguridad Nuclear y Salvaguardias, Central Nucleoelectrica de Laguna Verde (CNLV), Secretaria de Energia (Mexico) and Karlsruhe Institute of Technology (Germany) among others. This user group will be responsible for using the software and provide feedback to the development equipment in order that progress meets the needs of the regulator and industry; in this case the CNLV. Finally, in order to bridge the gap between similar developments globally, they will make use of the latest super computing technology to speed up calculation times. This work intends to present to national nuclear community the project, so a description of the proposed methodology is given, as well as the goals and objectives to be pursued for the development of the

  4. 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

  5. Method to generate the first design of the reload pattern to be used with the Presto-B code in the simulation of the CNLV U-1 reactor; Metodo para generar el primer diseno del patron de recarga a ser utilizado con el codigo Presto-B, en la simulacion del reactor de la CNLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J.L.; Cortes C, C.C. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1992-08-15

    This guide is applied for the reload pattern's formation for mirror symmetry of a core room and in accordance with the Control Cell core technique (of the english Control Cell Core - CCC) for the PRESTO-B code. (Author)

  6. Method by chromatography of gases for the determination of made up of alcoholic fermentation in pineapple fruits (Ananas comosus [L.] Merr); Metodo por cromatografia de gases para la determinacion de compuestos de fermentacion alcoholica en frutos de pina (Ananas comosus [L.] Merr)

    Energy Technology Data Exchange (ETDEWEB)

    Murillo Williams, A

    2001-07-01

    The pineapple (Ananas comosus) it is used in the entire world for the fresh consumption or for processed products (canned, frozen, dehydrated). it is cultivated in a wide range of countries and in extreme latitudes.The factors of quality include: maturity, stability, size uniformity, absence of microbial deterioration, absence of burns for the sun, absence of blows, damage for insects and breaking, crowns, color, longitude and integrity. The factors that determine the longevity of a product can be physiologic or pathological. The physiologic conditions refer to the processes of degradation of the fabrics after the crop, while the pathological ones involve the attack of mushrooms and bacteria. In the case of the pineapple, a physiologic problem exists, called black heart, internal brewing (IB), brown endogenic stain ({sup m}ancha cafe endogena{sup ,} MCE) or chilling injury (CI) that can happen in any part of the world where it is cultivated. This problem has associated to the exhibition from the pineapple to low temperatures, so it is a challenge to manage a fruit like the pineapple that it cannot tolerate low temperatures without problems. In studies made about the physiologic changes that happen during the storage in controlled atmospheres in fruits, it has been observed that the ethanol and the acetaldehyde are volatile compounds associated with metabolic post-crop changes and that they have implication in the quality of the product. (author) [Spanish] La pina (Ananas comosus) se utiliza en todo el mundo para el consumo fresco o para productos procesados (enlatados, congelados, deshidratados). Se cultiva en un amplio rango de paises y en latitudes extremas. Los factores de calidad incluyen: madurez, firmeza, uniformidad de tamano, ausencia de deterioro microbiano, ausencia de quemaduras por el sol, ausencia de golpes, dano por insectos y quebraduras, corona, color, longitud e integridad. Los factores que determinan la longevidad de un producto pueden ser

  7. Criticality Studies in a Pilot Plant for Processing MTR-Type Irradiated Fuels; Estudios de Criticidad de una Planta Piloto para el Tratamiento de Combustibles Irradiados Tipo ' MTR '

    Energy Technology Data Exchange (ETDEWEB)

    Pereira Sanchez, G.; Uriarte Hueda, A. [Junta de Energia Nuclear, Division de Materiales Madrid (Spain)

    1966-05-15

    A number of theoretical studies on nuclear safety have been carried out in a pilot plant being constructed at the Junta de Energia Nuclear in Madrid for processing irradiated fuels from the MTR-type experimental reactor JEN-1. The study was carried out working with aqueous and organic solutions at two levels of {sup 235}U enrichment - 20% and 93%. The paper is divided into two main parts: the first deals with the individual items of equipment, and the interactions between these are studied in the second part. The calculations in this second part have been made using three different methods to make it more certain that the system as a whole can never be critical. The first method employed is based on the solid angle concept and makes it possible to fix the maximum {sup 235}U concentrations within the system. The second method, based on the albedo, supplies the value of the multiplication factor K of the whole assembly as a function of the concentration of {sup 235}U. In the last part, the distribution of the equipment is compared with other similar systems and experimental tests from other sources. Finally, the paper specifies the conditions for working the installation which ensure that a nuclear accident can never occur. (author) [Spanish] Se ha efectuado una serie de estudios teoricos sobre la seguridad nuclear de una planta piloto, que se encuentra en construccion en la Junta de Energfa Nuclear situada en Madrid, para el tratamiento de combustibles irradiados procedentes del reactor experimental JEN-1 del tipo MTR. El estudio se ha realizado utilizando disoluciones, tanto acuosas como organicas, con dos grados de enriquecimiento, 20% y 93% en {sup 235}U. Este trabajo comprende dos partes principales: en la primera se han considerado las distintas unidades del equipo individualmente y en la segunda se han estudiado las interacciones entre ellas. El calculo de esta segunda parte se ha hecho por tres metodos diferentes para tener una mayor seguridad de que el

  8. Revisiting the reactor pressure vessel for long-time operation; Revisitando la vasija a presion del reactor para largos tiempos de operacion

    Energy Technology Data Exchange (ETDEWEB)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-07-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIFFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  9. Technologies for tritium control in fission reactors moderated with heavy water; Tecnologias para control de tritio en reactores de fision moderados con agua pesada

    Energy Technology Data Exchange (ETDEWEB)

    Ramilo, L B; Gomez de Soler, S M [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors` belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the `on-line` detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs.

  10. Development and testing of a recorder and controller for a microalgae culture reactor; Desarrollo y prueba de un registrador y controlador para un reactor de cultivo de microalgas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel, Wilson; Reyes, Jose Fernando; Bruijn, Johannes; Hernandez, Alejandro [Universidad de Concepcion, Chilan (Chile). Facultad de Ingenieria Agricola. Dept. de Mecanizacion y Energia], Emails: wesquive@udec.cl., jreyes@udec.cl., jdebruij@udec.cl., alehernandez@udec.cl

    2010-07-01

    An electronic system to monitor and control operational variables in a Raceway type of reactor for the culture of the Scenedesmus spinosus microalgae and later production for biodiesel and mitigating CO{sub 2} was developed and tested. The electronic system is constituted by a micro controller, a card reader SD, a card SD, a real-time clock, a power supply, a screen GLCD, a keyboard and a card for data acquisition, all implemented for 4-20 mA and 0-5 V output sensors. Temperature, pH, electrical conductivity, dissolved oxygen and solar radiation were measured digitalized and saved every 10 minutes. These variables were digitalized and kept in the SD memory every 10 minutes. It was determined that the most favorable conditions for the proliferation of the culture are near pH neutral and a temperature of 30 deg C, existing a strong correlation between pH and the dissolved CO{sub 2} level. Using the digital outputs of temperature and pH of the microcontroller, the CO{sub 2} injection and the elimination of O{sub 2} were controlled to maintain an adequate environment for the development of the culture. (author)

  11. Non-Destructive Testing Methods Applied to Multi-Finned SAP Tubing for Nuclear-Fuel Elements; Essais Non Destructifs de Gaines a Ailettes, en Poudre d'Aluminium Frittee, pour Elements Combustibles; Nedestruktivnye metody ispytaniya rebristykh trub iz spechennogo alyuminikiog'o poroshka dlya yadernykh toplivnykh ehlementov; Metodos de Ensayo No Destructivo Aplicados a Tubos de SAP con Aletas Multiples Destinados a Elementos Combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Lund, S. A. [Danish Central Welding Institution, Copenhagen (Denmark); Knudsen, P. [Danish Atomic Energy Commission, Research Establishment, Risoe (Denmark)

    1965-09-15

    Energia Atomica de Dinamarca ha emprendido el estudio de un reactor de potencia con refrigerante organico y moderador de agua pesada. Los correspondientes elementos combustibles consisten en haces de 19 barras formadas por pastillas de dioxido de uranio sinterizado, encerradas en tubos de producto de aluminio sinterizado (SAP), de 2 m de longitud, provistos de aletas helicoidales. Para obtener condiciones optimas de transmision de calor y mantener la integridad del elemento combustible durante el funcionamiento del reactor, es necesario contar con tubos de muy alta calidad. Se citan dos ejemplos que ponen de manifiesto las estrechas tolerancias dimensionales adoptadas. Para asegurar una calidad adecuada de los tubos, se establecio un control de calidad muy estricto, basado en gran medida en la aplicacion de metodos no destructivos. Se describen en esta memoria las tecnicas desarrolladas para medir el espesor de pared y los diametros, y para descubrir defectos. La compleja seccion transversal, con 24 aletas, impide aplicar metodos ultrasonicos o de comentes de Foucault para medir el espesor de la pared. Por consiguiente, se desarrollo un calibre registrador de rayos beta, cuyo funcionamiento se basa en la atenuacion sufrida por la radiacion beta proveniente de una fuente de {sup 90}Sr colocada en el interior del tubo. Para el registro continuo del espesor de la pared del tubo con seccion transversal mas simple, de 12 aletas, se utiliza un metodo ultrasonico de resonancia por inmersion. Los diametros interno y externo (entre puntos de aletas) se registran de manera continua mediante calibres neumaticos rapidos. Las fallas se detectan mediante la tecnica de eco de impulsos ultrasonicos, y examinando los tubos con corrientes de Foucault. El metodo ultrasonico permite descubrir facilmente las fisuras transversales, pero hasta ahora ha sido imposible utilizarlo para la deteccion de defectos longitudinales. Por consiguiente, ademas del ensayo ultrasonico, se aplica el examen con

  12. 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...

  13. 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

  14. 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.

  15. Ultrasonic Water-Gap Measurements in MTR Fuel Elements; Mesure par Ultrasons des Espaces Intercalaires dans les Elements Combustibles des Reacteurs d'Essai de Materiaux; Izmereniya vodyanogo zazora v teplovydelyayushchikh ehlementakh dlya materialovedcheskogo reaktora s pomoshch'yu ul'trazvuka; Medicion Ultrasonica de la Capa de Agua en Elementos Combustibles para Reactores de Ensayo de Materiales

    Energy Technology Data Exchange (ETDEWEB)

    Deknock, R. [Metallurgy Department, S.C.K./C.E.N., Mol (Belgium)

    1965-10-15

    distance intercalaire fournit a un enregistreur une tension stable de sortie de 1 V. Il est facile de mesurer les variations des distances intercalaires avec une precision de 5 {mu}m. Les mesures ont ete faites pour plusieurs elements combustibles. Les resultats et la reproductibilite sont tres satisfaisants. (author) [Spanish] Los elevados flujos termicos que suelen alcanzarse en los recientes reactores de ensayo de materiales, exigen recorridos adecuados para lograr una transmisiun uniforme de calor y una disipacion segura del mismo, evitando asf la formacion de vapor en la masa. Ademas, a fin de controlar el hinchamiento y el comportamiento del combustible en el reactor, tambien debe medirse la capa de agua en experimentos realizados despues de la irradiacion, con elementos combustibles agotados. A tal efecto se ha disenado una sonda ultrasonica destinada a medir, en una longitud de 1 m el espesor de 3 mm de agua correspondiente al elemento combustible BR-2. En el caso de los experimentos posteriores a la irradiacion, es necesario trabajar con el elemento combustible sumergido en un tanque de agua, a profundidad no menor de 6 m. La sonda puede resistir una prolongada inmersion en agua, y no le afectan las dosis normales de radiacion gamma. Aunque proyectado conforme al metodo usual de reflexion de impulsos, el sistema permite separar pulsos emitidos y reflejados, usando un cristal ferro-electrico de 10 MHz, con elevada disipacion inherente de energia. Puede usarse un osciioscopio para la lectura, en cuyo caso el tiempo se representa en el eje horizontal, regulandose la velocidad de barrido de manera que sea directamente proporcional a la velocidad de propagacion de la onda, es decir, al espesor de la capa de agua. Este tipo de representacion da resultados satisfactorios cuando setrata de un numero limitado de mediciones, pero sin duda resulta mas conveniente el registro grafico. En este caso, se da a los impulsos emitidos y reflejados la forma deseada y se les inyecta

  16. A {beta} - {gamma} coincidence; Metodo de coincidencias {beta} - {gamma}

    Energy Technology Data Exchange (ETDEWEB)

    Agullo, F

    1960-07-01

    A {beta} - {gamma} coincidence method for absolute counting is given. The fundamental principles are revised and the experimental part is detailed. The results from {sup 1}98 Au irradiated in the JEN 1 Swimming pool reactor are given. The maximal accuracy is 1 per cent. (Author) 11 refs.

  17. Alternate method for to realize image fusion; Metodo alterno para realizar fusion de imagenes

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, L; Hernandez, F; Fernandez, R [Departamento de Medicina Nuclear, Imagenologia Diagnostica. Centro Medico de Xalapa, Veracruz (Mexico)

    2005-07-01

    At the present time the image departments have the necessity of carrying out image fusion obtained of diverse apparatuses. Conventionally its fuse resonance or tomography images by X-rays with functional images as the gammagrams and PET images. The fusion technology is for sale with the modern image equipment and not all the cabinets of nuclear medicine have access to it. By this reason we analyze, study and we find a solution so that all the cabinets of nuclear medicine can benefit of the image fusion. The first indispensable requirement is to have a personal computer with capacity to put up image digitizer cards. It is also possible, if one has a gamma camera that can export images in JPG, GIF, TIFF or BMP formats, to do without of the digitizer card and to record the images in a disk to be able to use them in the personal computer. It is required of one of the following commercially available graph design programs: Corel Draw, Photo Shop, FreeHand, Illustrator or Macromedia Flash that are those that we evaluate and that its allow to make the images fusion. Anyone of them works well and a short training is required to be able to manage them. It is necessary a photographic digital camera with a resolution of at least 3.0 mega pixel. The procedure consists on taking photographic images of the radiological studies that the patient already has, selecting those demonstrative images of the pathology in study and that its can also be concordant with the images that we have created in the gammagraphic studies, whether for planar or tomographic. We transfer the images to the personal computer and we read them with the graph design program. To continuation also reads the gammagraphic images. We use those digital tools to make transparent the images, to clip them, to adjust the sizes and to create the fused images. The process is manual and it is requires of ability and experience to choose the images, the cuts, those sizes and the transparency grade. (Author)

  18. New method for protection of parallel generator; Novo metodo para protecao do gerador em paralelismo

    Energy Technology Data Exchange (ETDEWEB)

    Silva, M R.C. da [Elfa-Seg Eletronica Ltda. (Brazil)

    1988-07-01

    The protection of synchronous machinery, especially generators working in parallel with the pertaining electric power utility have been extensively discussed specially because of the growing importance of co-generation in Brazil. This work discusses existing efficient methods and suggests new ways of proceeding this protection. 8 refs., 2 figs.

  19. Geostatistical methods for the integrated information; Metodos geoestadisticos para la integracion de informacion

    Energy Technology Data Exchange (ETDEWEB)

    Cassiraga, E F; Gomez-Hernandez, J J [Departamento de Ingenieria Hidraulica y Medio Ambiente, Universidad Politecnica de Valencia, Valencia (Spain)

    1996-10-01

    The main objective of this report is to describe the different geostatistical techniques to use the geophysical and hydrological parameters. We analyze the characteristics of estimation methods used in others studies.

  20. CONSERVACION DE MANZANAS GRANNY SMITH MINIMAMENTE PROCESADAS POR METODOS COMBINADOS

    OpenAIRE

    GALLEGUILLOS CANALES, PAMELA CRISTINA; GALLEGUILLOS CANALES, PAMELA CRISTINA

    2011-01-01

    Se ha demostrado que el consumo de frutas y hortalizas tiene un efecto beneficioso para la salud. Debido a su alto contenido de antioxidantes, constituye una de las maneras más efectivas para reducir el riesgo de enfermedades crónicas no transmisibles (Speisky, 2006; Block et al., 1992; Knekt et al., 1997; Ness y Powles, 1997; Doll, 1990; Wang et al., 1996; Del Caro et al., 2004). El aumento de muertes por enfermedades como cáncer, diabetes y enfermedades cardiovasculares, ha llevado a ent...

  1. Avaliação dos métodos de amostragem para fauna perifítica em macrófitas na Reserva da Biosfera, Serra do Espinhaço, Estado de Minas Gerais, Brasil - DOI: 10.4025/actascibiolsci.v30i3.377 Evaluation of sampling methods for periphytic fauna in macrophytes at the Espinhaço Mountain Range Biosphere Reserve, Minas Gerais State, Brazil - DOI: 10.4025/actascibiolsci.v30i3.377

    Directory of Open Access Journals (Sweden)

    Cristiane Machado López

    2008-10-01

    Full Text Available Os métodos “Jarra”, “Remocao Manual” e “Draga de Eckman modificada” foram avaliados para amostrar a fauna perifitica associada a macrofitas aquaticas. Foram coletadas 63 amostras em cinco ambientes lenticos e tres loticos na reserva da biosfera da Serra do Espinhaco (Estado de Minas Gerais, Brasil. Os testes estatisticos Anova e Tukey foram feitos para riqueza de Protista, Rotifera e Crustacea, enquanto para a abundancia de Protista, Rotifera, Crustacea, Gastrotricha, Tardigrada e Nematoda foram avaliados os percentuais. Os protozoarios e rotiferos representaram 80% da abundancia e riqueza da comunidade. Nos ecossistemas avaliados todos os metodos foram relevantes para Protista, por outro lado, o metodo da Jarra foi o mais adequado para a analise de Crustacea. Entre os metodos, a Draga foi menos indicada para os grupos de microinvertebrados nos ecossistemas aquaticos. Os metodos Remocao Manual e Draga foram apropriados para analisar Rotifera. A abundancia de Gastrotricha e Tardigrada demonstrou melhores resultados pelo metodo da Jarra e Nematoda pelo metodo da Draga. Os tres metodos sao apropriados para amostragem da fauna perifitica em ambos os sistemas aquaticos. Entretanto, e importante estar ciente de que para cada tipo de ecossistema a amostragem da comunidade faunistica requer um metodo especifico para obter a melhor performance.The methods “Jar”, “Manual Removal” and “modified Ekman`s Dredge” were evaluated for sampling periphyton fauna associated to aquatic macrophytes. Sixty three samples were collected from five lentic and three lotic water bodies at Espinhaço Mountain Range Biosphere Reserve (Minas Gerais, Brazil. ANOVA and Tukey statistical tests were performed for Protista, Rotifera and Crustacea richness, whilst Protista, Rotifera, Crustacea, Gastrotricha, Tardigrada and Nematoda, abundance were evaluate by percentage. Amongst the three methods, Dregde is less indicated for different water bodies systems

  2. Development of an analytic outline for the aflatoxins analysis in grains and flours; Desarrollo de un esquema analitico para el analisis de aflatoxinas en granos y harinas

    Energy Technology Data Exchange (ETDEWEB)

    Sibaja Adams, Roxana

    2000-07-01

    The instrumental and analytic conditions were optimized for the aflatoxine determination B1, B2, 1 and G2 in corn and peanut byl iquid chromatography of high discharge following the analyzing method AOAC 994,08. Besides, it was defined a function for evaluating the dependence of the chromatographic discharge with the aflatoxine concentration. The analyzing method was validated, and four calibration curves were obtained for the aflatoxine B1, B2, G1 and G2, which turned to have a heterocedastico behavior. The applicability of this method was demonstrated, obtaining imagines of appropriate merit and comparable with those reported by the AOAC. Additionally, the applicability of the chromatographic method was demonstrated in fine layer for the presumptive analysis of aflatoxine, allowing both methods to propose an outline of reliable analysis of real samples. [Spanish] Se optimizaron las condiciones instrumentales y analiticas para la determinacion de aflatoxinas B1, B2, G1 y G2 en maiz y mani por Cromatografia liquida de alto desempeno siguiendo el metodo de analisis AOAC 994,08. Ademas, se definio una funcion para evaluar la dependencia del desempeno cromatografico con la concentracion de aflatoxinas.Se valido el metodo de analisis y se obtuvieron las cuatro curvas de calibracion para las aflatoxinas B1, B2, G1 y G2, las que resultaron tener un comportamiento heterocedastico. Se demostro la aplicabilidad del metodo, obteniendose figuras de merito adecuadas y comparables con las reportadas por el AOAC.Adicionalmente, se demostro la aplicabilidad del metodo de cromatografia en capa fina para el analisis presuntivo de aflatoxinas, permitiendo ambos metodos proponer un esquema de analisis confiable de muestras reales.

  3. Matriz para ingenieria de tejido oseo: modificacion superficial con zinc Estudio preliminar

    OpenAIRE

    Martinez, Cristian; Olmedo, Daniel; Ozols, Andres

    2013-01-01

    El objetivo del presente trabajo es desarrollar un metodo de sintesis de matrices para ingenieria de tejidos, capaces de acelerar los procesos reparativos y disminuir el riesgo de infeccion, destinadas a cirugias reconstructivas que involucren el tejido oseo. A este fin se emplea la fase mineral de hueso bovino, la hidroxiapatita, conservando su estructura tridimensional. Esta se obtiene a partir de un proceso de lavado quimico, seguido de un proceso termico que elimina todos los componentes ...

  4. Módulos interactivos para el aprendizaje de la semántica.

    OpenAIRE

    Aquino Palacios, Ingrid

    2012-01-01

    Se proponen modulos interactivos para lograr el aprendizaje de la semantica en alumnos del V semestre de la Carrera de Lenguas, Literatura y Comunicacion de la Facultad de Educacion de la UNCP. Los metodos utilizados fueron el experimental y el bibliografico-documental. Los instrumentos utilizados fueron la prueba objetiva, prueba de ensayo y la prueba pedagogica. Se logró mejorar el aprendizaje de la semántica 

  5. USO DA SISTEMATIZAÇÃO DA ASSISTÊNCIA DE ENFERMAGEM (SAE: UMA FERRAMENTA PARA REALIZAÇÃO DA AUDITORIA DE QUALIDADE

    Directory of Open Access Journals (Sweden)

    Maria Izelta da Silva Santos

    2013-04-01

    Full Text Available Realizar uma revisao bibliografica acerca da Sistematização da Assistencia de Enfermagem (SAE como instrumento para auditoria de qualidade. Metodos: Foram revisados 12 artigos cientificos sobre auditoria e 10 artigos sobre SAE, bem como dois livros sobre os assuntos abordados. Esta revisao e de carater descritivo. Resultados: Por meio da pesquisa, tornou-se evidente a correlação da SAE com a auditoria. A SAE bem implementada pode contribuir para uma auditoria de qualidade.

  6. DOLOČITEV POVRŠINSKE NAPETOSTI Z METODO KAPILARNEGA DVIGA DVOFAZNIH SISTEMOV

    OpenAIRE

    Kravanja, Gregor

    2014-01-01

    Namen magistrske naloge je bil postavitev merilne naprave in razvoj nove metode merjenja površinske napetosti s kapilarnim dvigom dvofaznih sistemov v okolici kritične točke. Za pridobitev natančnih in primerljivih meritev je bilo potrebno poznati natančni notranji premer tankih kapilar. Določili smo ga z metodo laserskega tipanja na nemški koordinatni merilni napravi ZEISS tipa UMC-850 s pomočjo merilne programske opreme CALYPSO 5.1.4. Za merjenje ravnotežne višine smo uporabili računalniški...

  7. Ensenanza de la Astronomia a Traves de Metodos no Tradicionales

    Science.gov (United States)

    Tignanelli, H. L.

    1990-11-01

    REUMEN: Se presentan los aspects pri nc ipales de a ense? anza de la astronor:. a para -; s. En esta cc .unicaci #n, 5 ha especial mfasis em Ia descripci':"n de las caracteristicas y las posibi lidades peda gicas de los no tradicionales de aprendiZaje. E' : In the following the principal aspects of teaching of astrono ..y for children) are oresented. In this paper, special emphasis has been given to desc rib the characteristics and pedagogical possibilities of the non traditional methods of learning. : TEACHING

  8. Impacto orcamentario da utilizacao do Metodo Canguru no cuidado neonatal

    Directory of Open Access Journals (Sweden)

    Aline Piovezan Entringer

    2013-10-01

    Full Text Available OBJETIVO Estimar o impacto orçamentário da utilização do Método Canguru na rede municipal de saúde. MÉTODOS Um modelo de decisão analítico foi desenvolvido para simular os custos do Método Canguru e Unidade Intermediária Neonatal no Rio de Janeiro, RJ, em 2011. A população de referência foi constituída pelos recém-nascidos estáveis clinicamente, que podem receber assistência nas duas modalidades de cuidado. O impacto orçamentário foi estimado para uma coorte hipotética de 1.000 recém-nascidos elegíveis em um ano. A proporção de recém-nascidos elegíveis que recebem assistência nas duas modalidades foi obtida por coleta de dados nas maternidades incluídas no estudo. As probabilidades dos eventos e o consumo de recursos de saúde, no período da assistência, foram incorporados ao modelo. Cenários foram desenvolvidos para refletir a adoção do método Canguru em maior ou menor escala. RESULTADOS A utilização do Método Canguru significou redução de gastos equivalente a 16% em um ano, se todos os recém-nascidos elegíveis fossem assistidos por esse método. CONCLUSÕES A opção Método Canguru é de menor custo comparado com a da Unidade Intermediária Neonatal. A análise de impacto orçamentário da utilização desse método no Sistema Único de Saúde indicou economia importante para o período de um ano.

  9. Analytic techniques to quantify Tetrachlorohydroquinone and Chloranil; Tecnicas analiticas para cuantificar Tetraclorohidroquinona y Cloranilo

    Energy Technology Data Exchange (ETDEWEB)

    Castillo Escobedo, Ma. Teresa; Gutierrez de Gonzalez, Luz Ma; Gojon Zorrilla, Gabriel [Universidad Autonoma de Nuevo Leon, Monterrey, Nuevo Leon (Mexico)

    1995-02-01

    Tetrachlorohydroquinone (TCHQ) was determined by two methods: iodometry and cerimetry. Tritation with aqueous ceric ammonium sulphate using potentiometric end-point detection proved to be the method of choice on account of its coefficient of variation (CV=0.475%) and its excellent accuracy. Chloranil (TCQ) was quantified by titration with aqueous ascorbic acid in acetone-water-hexamethylene tetramine (HMT), the golden-yellow color of the TCQ-HMT complex disappearing at the end point. This method is accurate and has CV=0.396%. [Spanish] Se cuantifico la tetraclorohidroquinona (TCHQ) mediante una tecnica yodometrica y un metodo potenciometrico, basado en el uso de sulfato cerico amoniacal. Se concluyo que el metodo potenciometrico es el mejor, ya que tiene un coeficiente de variacion (CV) de 0.475%, y una exactitud muy aceptable. Se desarrollo un tecnica para cuantificar el Cloranilo (TCQ) por titulacion con Acido Ascorbico en presencia de Hexametilentetramina (HMT) en medio acetona-agua; el punto final se determino por la separacion del color amarillo oro del complejo TCQ-HMT. Esta tecnica presenta un coeficiente de variacion (CV) de 0.396% y una exactitud aceptable. Se realizaron pruebas de hipotesis para verificar la exactitud de los metodos elegidos.

  10. Techniques for contact and contact with friction problems; Tecnicas para problemas de contacto y contacto con friccion

    Energy Technology Data Exchange (ETDEWEB)

    Velandia Arana, Gonzalo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1990-12-31

    Different numerical techniques are presented based in the finite element method to obtain numerical solutions to contact and contact with friction problems between solid bodies, and compared between each other. [Espanol] Se presentan diferentes tecnicas numericas basadas en el metodo de elementos finitos para la obtencion de soluciones numericas de problemas de contacto y contacto con friccion entre cuerpos solidos, y se comparan entre si.

  11. Techniques for contact and contact with friction problems; Tecnicas para problemas de contacto y contacto con friccion

    Energy Technology Data Exchange (ETDEWEB)

    Velandia Arana, Gonzalo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1989-12-31

    Different numerical techniques are presented based in the finite element method to obtain numerical solutions to contact and contact with friction problems between solid bodies, and compared between each other. [Espanol] Se presentan diferentes tecnicas numericas basadas en el metodo de elementos finitos para la obtencion de soluciones numericas de problemas de contacto y contacto con friccion entre cuerpos solidos, y se comparan entre si.

  12. 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).

  13. 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.

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. 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...

  19. High-Temperature Gas-Cooled Reactor Critical Experiment and its Application to Thorium Absorption Rates; Experience Critique pour l'Etude d'un Reacteur a Haute Temperature, Refroidi par un Gaz et son Application a la Determination des Taux d'Absorption du Thorium; Kriticheskij opyt, postavlennyj na vysokotemperaturnom reaktore s gazovym okhlazhdeniem, i primenenie ego dlya opredeleniya stepeni pogloshcheniya toriya; Experimento Critico Efectuado en un Reactor de Elevada Temperatura Refrigerado por Gas y su Aplicacion para Calcular los Indices de Absorcion del Torio

    Energy Technology Data Exchange (ETDEWEB)

    Bardes, R. G.; Brown, J. R.; Drake, M. K.; Fischer, P. U.; Pound, D. C.; Sampson, J. B.; Stewart, H. B. [General Dynamics Corporation,San Diego, CA (United States)

    1964-04-15

    variations en fonction de la temperature (coefficient Doppler d'activation). Pour mesurer le coefficient Doppler par oscillation de la reactivite, on se sert de la totalite de la cartouche centrale au cours d'une operation qui permet de la porter a une temperature pouvant atteindre 425 Degree-Sign C et d'eliminer experimentalement les effets qui ne contribuent pas a la capture de neutrons de resonance par le thorium. Pour toute une gamme d'experiences, les resultats obtenus concordent bien avec les resultats theoriques. Le memoire decrit la mesure de l'integrale de resonance pour le thorium et ses variations en fonction de la temperature. Dans le procede que l'on a mis au point pour mesurer la capture de resonance, on utilise l'or comme etalon et le vanadium comme matiere donnant le taux d'absorption en 1/v. Ce procede a ete choisi parce que le thorium est disperse dans le graphite et qu'il est difficile d'appliquer le procede habituel du rapport cadmium. Les resultats empiriques concordent bien avec les resultats theoriques dans une large gamme de variables. En outre, les resultats des mesures du coefficient Doppler par les deux methodes (oscillation de la reactivite et activation) concordent. Les auteurs estiment que ce fait merite d'etre releve, car dans les ouvrages publies jusqu'ici ces deux procedes donnaient des resultats differents. (author) [Spanish] Al definir los principios teoricos del reactor de elevada temperatura refrigerado por gas, y de su primer prototipo en Peach Bottom, la General Atomic decidio efectuar un experimento critico con miras a reunir ciertos datos de entrada requeridos para el analisis nuclear. Debido a las necesidades especificas de la teoria de las construcciones nucleares, en lo que atane a los datos de entrada acerca de la absorcion del torio, los autores elaboraron un sistema experimental formado por un conjunto critico con reticulado central rodeado por una region amortiguadora y otra activado'. Este tipo de conjunto, en el que el

  20. Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2015-07-01

    Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

  1. Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor; Diseno y construccion de un sistema electronico automatico de medicion y graficado del flujo neutronico para el reactor subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

  2. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  3. Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor; Desarrollo de un sistema de vigilancia remota, por medio del monitoreo de parametros claves de seguridad, para un reactor nuclear de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Urcia, Agustin; Arrieta, Rolando [Direccion de Produccion, Instituto Peruano de Energia Nuclear, Lima (Peru); Baltuano, Oscar; Chan, Renzo [Direccion de Investigacion y Desarrollo, Instituto Peruano de Energia Nuclear, Lima (Peru); Tincopa, Jean Pierre [Facultad de Ingenieria Electrica y Electronica, Universidad Nacional del Callao, Callao (Peru); Urquizo, Rafael [Facultad de Ingenieria Electronica, Universidad Tecnologica del Peru, Lima (Peru); Rosas, Bernick [Facultad de Ingenieria Electronica, Universidad Nacional de Ingenieria, Lima (Peru)

    2014-07-01

    This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

  4. De "átomos para la paz" a los reactores de potencia: Tecnología y política nuclear en la Argentina (1955-1976)

    OpenAIRE

    Hurtado de Mendoza, Diego

    2005-01-01

    Durante el período 1955-76, el programa nuclear argentino se integró a la arena internacional; su Comisión Nacional de Energía Atómica construyó cuatro reactores de investigación, adquirió a una empresa alemana y puso en marcha el primer reactor de potencia Atucha I, y compró a una empresa canadiense un segundo reactor de potencia. En este artículo se examinan estos desarrollos en relación con el contexto político local y con el panorama nuclear internacional. En particular, se analizan la po...

  5. 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)

  6. 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).

  7. Some Non-Destructive Testing Methods Applicable to Sintered Materials; Quelques Methodes d'Essais Non Destructifs Applicables aux Materiaux Frittes; Nekotorye metody nedestruktivnykh ispytanii, primenimye k spechennym materialam; Algunos Metodos de Ensayo No Destructivo Aplicables a los Materiales Sinterizados

    Energy Technology Data Exchange (ETDEWEB)

    Labusca, Elena; Mirion, I.; Andreescu, N.; Alecu, M.; Biscoveanu, I. [Institut de Physique Atomique, Bucarest (Romania)

    1965-10-15

    solides frittes, a l'aide de la metallographie et de la microscopie electronique. Ces methodes mettent en evidence l'homogeneite de la structure, les dimensions et l'orientation des grains, la presence de defauts divers (inclusions, pores) et revolution meme du processus de frittage, y compris la formation des cristaux, la croissance granulaire, etc. Dans quelques cas, on peut combiner l'examen microscopique a des essais de microdurete. Cet examen de la structure microcristalline represente l'une des methodes principales de controle qualitatif des materiaux frittes, et ne peut etre remplace par aucun autre moyen d'investigation. 2. Controle du degre de consolidation, qui determine essentiellement la qualite des materiaux frittes. Ce controle est effectue par la mesure de quelques proprietes, telles que la conductibilite electrique et thermique, en correlation avec la densite, etant donne que la conductibilite des materiaux frittes est directement porportionnelle au degre de frittage. On a essaye aussi une methode adequate de controle de la porosite; on a obtenu des donnees experimentales interessantes, surtout au point de vue de la porosite libre, laquelle est susceptible aux inclusions gazeuses. Le memoire contient des donnees experimentales concernant l'application de ces methodes de controle a quelques materiaux frittes interessants pour la technologie nucleaire. (author) [Spanish] Teniendo en cuenta la estructura granular especifica de los materiales sinterizados, elaborados a partir de polvos, cuyo proceso de consolidacion se desarrolla en funcion del tratamiento de sinterizacion, los autores han estudiado algunos metodos para verificar el grado de sinterizacion y controlar ciertas propiedades. Entre los metodos no destructivos utilizados, se mencionan en la presente memoria: 1. Examen de la estructura cristalina de los solidos sinterizados por metalografia y microscopfa electronica. Estos metodos ponen de manifiesto el giado de homogeneidad estructural, la dimension

  8. Osservazioni sul principio di legalità come idea e come metodo nell’esperienza giuridica della Chiesa

    Directory of Open Access Journals (Sweden)

    Beatrice Serra

    2012-10-01

    SOMMARIO: 1. Introduzione - 2. – La struttura essenziale del concetto di legalità e la sua realizzazione radicale nel diritto della Chiesa. - 3. L’idea di legalità come “regola che sta prima” nel ius commune e la sua coesistenza con un approccio empirico, essenzialmente giurisprudenziale e dottrinale, di costruzione del diritto. Il principio di legalità come metodo di produzione del diritto nel pensiero giuridico moderno- 4. Il legame storico-concettuale fra il principio di legalità e i Codici moderni - 5. Il Codex iuris canonici del 1917 e il principio di legalità come metodo di costruzione del ius ecclesiae in funzione della certezza del diritto - 6. (segueUlteriori riflessi della prima codificazione canonica sul principio di legalità.:

  9. 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)

  10. Metodologia de la Administración Federal de Aviación para el Diseño de estructuras de Pavimento flexible para aeropuertos

    Directory of Open Access Journals (Sweden)

    Julián Rodrigo Quintero González

    2013-08-01

    Full Text Available Se representan los aspectos técnicos fundamentales y principalesconsideraciones de la metodologia propuesta por la Administracion Federal de Aviación FAA para el diseño de estructuras de pavimento para aeropuertos. Se exponen los aspectos sobre las variables consideradas en el diseño (estructurales tránsito, factores ambientales y factores intrinsecos, se establecen las caracteristicas de las variables consideradas por el metodo (subrasante y materiales para la estructura, geometría del tren de aterrizaje, número de decolajes equivalentes y peso bruto maximo del decolaje y se explica el procedimiento utilizado para el tratamiento de la información requerida en diseño estructural de pavimentos flexibles para aeropuertos.

  11. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  12. Computerized supervision and control system for movement at the RP-10 reactor control rods bank; Sistema de supervicion y mando computarizado para el movimiento en banco de las barras de control del Reactor RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Padilla M, C E

    1998-07-01

    The project involves the use of a compatible microcomputer, Labwindows/CVI software, as well as National Instruments data acquisition cards AT-MIO16-E10 and PC-DIO96 to modify the sequence of movement of the reactor's rods and control them from a graphic interface in a computer's monitor. This graphic presentation is set as console of virtual instruments from where rod movement can be conducted. Normal rod movement, bank rod movement, and rod calibration have been considered. These experiences involve different logic of rod movements, which will determine movement sequence. Control of the automatic range of a current amplifier module was also considered. This module is know as 'automatic pilot amplifier' and given the strategic location of its detector (compensated ionizing camera) at the reactor's core, it delivers neutron flux current considered as reference to superficial neutron flux distribution at the reactor's core. Lecture and monitoring of this signal allows taking the reactor to a certain power, current of this signal is proportional to the power we want the reactor to reach. Advantages obtained with this system include the update of the control console, more uniform distribution of neutron flux, with lower and uniform burnup of nuclear fuel. (author)

  13. Actions to reduce radioactive emissions: prevention of containment failure by flooding Containment and Reactor Cavity; Acciones para la reduccion de emisiones radiactivas: prevencion del fallo de la Contencion mediante la inundacion de la Contencion y de la Cavidad del Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fornos Herrando, J.

    2013-07-01

    The reactor cavity of Asco and Vandellos II is dry type, thus a severe accident leading to vessel failure might potentially end up resulting in the loss of containment integrity, depending on the viability to cool the molten core. Therefore, significant radioactive emissions could be released to outside. In the framework of Fukushima Stress Tests, ANAV has analyzed the convenience of carrying out different actions to prevent failure of the containment integrity in order to reduce radioactive emissions. The aim of this paper is to present and describe the main phenomenological aspects associated with two of these actions: containment flooding and reactor cavity flooding.

  14. Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor; Desarrollo de un simulador para diseno y prueba de controladores de potencia en un reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Perez M, C.; Benitez R, J.S.; Lopez C, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

  15. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors; Calificacion del programa WIMS de calculo neutronico para diseno, seguimiento de operacion y analisis de accidentes de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M [Ente Nacional Regulador Nuclear, Buenos Aires (Argentina)

    1997-12-31

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

  16. Critical Consideration of the Methods of Calculation Used in the Evaluation of the Absorbed Dose to the Skin in Cases of External Contamination; Observations Critiques sur les Methodes de Calcul Utilisees pour l'Evaluacion de la Dose Absorbee par la Peau en Cas de Contamination Externe; 041a 0420 0414 ; Observaciones Criticas Sobre los Metodos de Calculo Utilizados para Evaluar la Dosis Absorbida por la Piel en Caso de Contaminacion Externa

    Energy Technology Data Exchange (ETDEWEB)

    Casnati, E.; Breuer, F. [Gruppo di Dosimetria e Standardizzazione, CNEN, Centro di Studi Nucleari della Casaccia, Rome (Italy)

    1965-06-15

    resultats permettent de conclure que, dans la geometrie consideree, on peut determiner par des formules de type exponentielle, d'une part la dose imputable aux rayons beta et, d'autre part, la dose imputable aux rayons gamma pour la couche basale de la peau;- on peut considerer les resultats comme acceptables du point de vue de la radioprotection. (author) [Spanish] La dosis absorbida, suele calcularse mediante la formula de Loevinger cuando se trate de particulas beta mientras que para las radiaciones electromagneticas se recurre habitualmente a una formula exponencial. En la memoria, los autores se proponen examinar si en el caso de contaminacion cutanea puede considerarse satisfactorio el empleo de una formula exponencial (la formula de Rossi y Ellis) y de una formula similar para calcular la dosis beta y la dosis gamma, respectivamente. Con este objeto, se han comparado los resultados.obtenidos para este tipo especial de geometria por medio de la formula de Rossi y Ellis y con la formula de Loevinger para partfculas beta; por lo que respecta a la radiacion gamma, se han comparado los valores obtenidos mediante la formula exponencial y con otras dos formulas mas complejas, que se ajustan a las condiciones de acumulacion fotonica maxima nula, respectivamente. Estos resultados permiten sacar la conclusion de que, dadas las condiciones geometricas de que se trata, pueden determinarse mediante la formula de tipo exponencial tanto la dosis de radiacion beta como la dosis de radiacion gamma en la dermis. Los resultados son aceptables desde el punto de vista de la proteccion radiologica. (author) [Russian] Obychno pri podschete dozy ispol'zuetsja formula Laevingera dlja chastic i jeksponencial'naja formula dlja jelektromagnitnyh izluchenij. V doklade rassmatrivaetsja vopros o tom, mozhno li schitat' udovletvoritel'nym ispol'zovanie jeksponencial'noj formuly (podobnoj formule Rossi i Jellisa) i analogichnyh formul pri podschete sootvetstvenno beta- i gamma-dozy. S jetoj cel

  17. Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Andrzejewski, Claudio de Sa

    2005-07-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  18. Development of a digital card to simulate period transients in research reactors; Desenvolvimento de um cartao digital para simulacao da variacao do periodo em reatores de pesquisa

    Energy Technology Data Exchange (ETDEWEB)

    Masotti, Paulo Henrique Ferraz

    1999-07-01

    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)

  19. Analysis of reactor power behaviour using estimation of period for the gain adaptation in a state feedback controller; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Benitez R, J.S. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Perez C, J.H. [CINVESTAV, IPN, A.P. 14740 07000 Mexico D.F. (Mexico); Rivero G, T. [ITT, 50140 Metepec, Estado de Mexico (Mexico)

    2008-07-01

    In this paper a novel procedure for power regulation in a TRIGA Mark III nuclear reactor is presented. The control scheme combines state variable feedback with a first order predictor, which is incorporated to speed up the power response of the reactor without exceeding the safety requirement imposed by the reactor period. The simulation results using the proposed control strategy attains different values of steady-state power from different values of initial power in short time, complying at all times with the safety restriction imposed on the reactor period. The predictor, derived from the theory of first order numerical integration, produces very good results during the ascent of power. These results include a fast response and independence of the wide variety of potential operating conditions something not easy and even impossible to obtain with other procedures. By using this control scheme, the reactor period is maintained within safety limits during the start up of the reactor, which is normally the operating condition where an occurrence of a period scram is common. However, the predictor can not be used when the power is reaching the desired power level because the instantaneous power increases far above the desired level. Thus, when the power increases above certain power level, the state feedback gain is set constant to a predefined value. This causes some oscillations that decrease in a few seconds. Afterwards, the power response smoothly approaches, with a small overshoot, the desired power. This constraint on the use of the predictor prevents the unbounded increase of the neutron power. The control law proposed requires all the system's state variables. Since only the neutron power is available, it is necessary the estimation of the non measurable states. The key issue of the existence of a solution to this problem has been previously considered. One of the conclusions is that the point kinetic equations are observable under certain restrictions

  20. Development of a simplified methodology for the isotopic determination of fuel spent in Light Water Reactors; Desarrollo de una metodologia simplificada para la determinacion isotopica del combustible gastado en reactores de agua ligera

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez N, H.; Francois L, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: hermilo@lairn.fi-b.unam.mx

    2005-07-01

    The present work presents a simplified methodology to quantify the isotopic content of the spent fuel of light water reactors; their application is it specific to the Laguna Verde Nucleo electric Central by means of a balance cycle of 18 months. The methodology is divided in two parts: the first one consists on the development of a model of a simplified cell, for the isotopic quantification of the irradiated fuel. With this model the burnt one is simulated 48,000 MWD/TU of the fuel in the core of the reactor, taking like base one fuel assemble type 10x10 and using a two-dimensional simulator for a fuel cell of a light water reactor (CPM-3). The second part of the methodology is based on the creation from an isotopic decay model through an algorithm in C++ (decay) to evaluate the amount, by decay of the radionuclides, after having been irradiated the fuel until the time in which the reprocessing is made. Finally the method used for the quantification of the kilograms of uranium and obtained plutonium of a normalized quantity (1000 kg) of fuel irradiated in a reactor is presented. These results will allow later on to make analysis of the final disposition of the irradiated fuel. (Author)

  1. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  2. Sargent-IV Project. Development of new methodologies for safety analysis of Generation IV reactors; Proyecto SARGEB-IV. Desarrollo de nuevas metodologias de analisis de seguridad para reactores de Generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Gallego, E.; Jimenez, G.

    2013-07-01

    The main result of this paper is the proposal for the addition of new ingredients in the safety analysis methodologies for Generation-IV reactors that integrates the features of probabilistic safety analysis within deterministic. This ensures a higher degree of integration between the classical deterministic and probabilistic methodologies.

  3. Reactors licensing: proposal of an integrated quality and environment regulatory structure for nuclear research reactors in Brazil; Licenciamento de reatores: proposta de uma estrutura regulatoria integrada com abordagem em qualidade e meio ambiente para reatores de pesquisa no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Serra, Reynaldo Cavalcanti

    2014-07-01

    A new integrated regulatory structure based on quality and integrated issues has been proposed to be implemented on the licensing process of nuclear research reactors in Brazil. The study starts with a literature review about the licensing process in several countries, all of them members of the International Atomic Energy Agency. After this phase it is performed a comparative study with the Brazilian licensing process to identify good practices (positive aspects), the gaps on it and to propose an approach of an integrated quality and environmental management system, in order to contribute with a new licensing process scheme in Brazil. The literature review considered the following research nuclear reactors: Jules-Horowitz and OSIRIS (France), Hanaro (Korea), Maples 1 and 2 (Canada), OPAL (Australia), Pallas (Holand), ETRR-2 (Egypt) and IEA-R1 (Brazil). The current nuclear research reactors licensing process in Brazil is conducted by two regulatory bodies: the Brazilian National Nuclear Energy Commission (CNEN) and the Brazilian Institute of Environment and Renewable Natural Resources (IBAMA). CNEN is responsible by nuclear issues, while IBAMA by environmental one. To support the study it was applied a questionnaire and interviews based on the current regulatory structure to four nuclear research reactors in Brazil. Nowadays, the nuclear research reactor’s licensing process, in Brazil, has six phases and the environmental licensing process has three phases. A correlation study among these phases leads to a proposal of a new quality and environmental integrated licensing structure with four harmonized phases, hence reducing potential delays in this process. (author)

  4. Design and construction of a live insulator washing system for transformers; Diseno y construccion de un sistema de lavado en vivo para los aisladores de transformadores

    Energy Technology Data Exchange (ETDEWEB)

    Lizama-Camara, Y.A. [Universidad Veracruzana, Veracruz (Mexico)]. E-mail: yahir_lizama@ieee.org; Mendieta-Antunez, J.A.; Blanco-Brisset, E. [Industrias IEM, Tlalnepantla, Estado de Mexico (Mexico)]. E-mail: unamanu@hotmail.com; Olivares Galvan, J.C.; Escarela-Perez, R. [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Mexico, D.F. (Mexico)]. E-mails: jolivare_1999@yahoo.com; r.escarela@ieee.org

    2012-04-15

    Through the electrical industry history there have been developments of different cleaning methods to avoid the insulators flashover s due to pollution. This paper describes the principal cleaning methods applicable to transformers insulators, emphasizing the high pressure fixed-type live insulator washing method, which was applied for cleaning the insulators of 900 MVA transformer bank of the Laguna Verde power plant localized at the state of Veracruz in Mexico. We propose a transformer insulator cleaning methodology, which identifies the main variables to take into account (the voltage level of the transformers, the pollution level of the insulators, determination of the optimal wash time, the amount of water, the optimal pressure of water jet, the maximum conductivity of the water and the wind velocity), reference values are given for these variables. In addition, we present an economic cost analysis when applying a method of this kind in an electric substation. [Spanish] A lo largo de la historia de la industria electrica se han desarrollado diferentes metodos de limpieza para evitar las fallas de los aisladores de los transformadores debido a la contaminacion. Este articulo describe los principales metodos de limpieza aplicables a los aisladores de transformadores, enfatizando el sistema de lavado en vivo tipo fijo con agua a alta presion, metodo que fue aplicado para realizar la limpieza de los aisladores en el banco de transformadores de 900 MVA de la central electrica Laguna Verde, ubicada en el estado de Veracruz, en Mexico. Se propone una metodologia para la limpieza de los aisladores de transformadores, donde se identifican las principales variables a tomar en cuenta (el nivel de tension de los transformadores, nivel de contaminacion de los aisladores, determinacion del tiempo optimo de lavado, cantidad de precipitacion de agua, presion optima del chorro de agua, maxima conductividad del agua y las velocidades de los vientos) y se dan valores de

  5. Ultra-high-speed oscillographic techniques; Techniques d'oscillographie ultra-rapide; Metody sverkhskorostnoj ostsillografii; Metodos oscilograficos ultrarrapidos

    Energy Technology Data Exchange (ETDEWEB)

    Abercrombie, S; Elphick, B; Foster, H [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

    1962-04-15

    100 {Omega}; e) Des resistances de complement a large bande, destinees a l'extraction des signaux de declenchement de la base de temps; / ) Des bases de temps fournissant des signaux de l'ordre de quelques nanosecondes et comportant des dispositifs de declenchement et de suppression du faisceau; g) Un appareil de prise de vues a grande ouverture. En conclusion, le memoire examine certaines applications de l'electronique nucleaire. (author) [Spanish] Despues de un resumen historico de los metodos de utilizacion de oscilografos para senales transitorias de alta velocidad, los autores examinan sus limites de aplicacion practica. En la memoria se pasa revista a los tubos de rayos catodicos apropiados para estos oscilografos, tomando en cuenta especialmente los tipos de deflexion por onda progresiva. En la memoria se estudian los siguientes componentes de un oscilografo completo para senales transitorias: a) Un tubo de rayos catodicos con deflexion por onda progresiva con conexion coaxial balanceada y una anchura de banda de 2 GHz; b) Un transformador inversor de fase con una impedancia caracteristica de 100 y 50 {Omega} y una anchura de banda de 1,5 kHz a 1,5 GHz; c) Resistencias de terminacion coaxial para frecuencias del orden de los GHz; d) Condensadores de acoplamiento compensados para usarlos en lineas coaxiales con frecuencias comprendidas entre 5 kHz y 3 GHz y 100 {Omega} de impedancia caracteristica; e) Atenuadores resistivos de banda ancha, para la extraccion de senales de disparo de base de tiempos; f) Bases de tiempo del orden de los nanosegundos, con dispositivos disparadores y supresores ; g) Una camara de gran apertura. Finalmente, los autores examinan las aplicaciones de estos oscilografos en la electronica nuclear. (author) [Russian] Posle kratkogo opisaniya ostsillograficheskogo metoda izucheniya sverkhskorostnykh perekhodnykh izluchenij obsuzhdayutsya prakticheskie vozmozhnosti ehtogo apparata. Daetsya obzor podkhodyashchikh katodnykh ehlektronnykh lamp

  6. Cleaning of porous filters in fluidized bed reactors. Use of one ejector for various filters; Limpieza de filtros porosos en reactores lecho fluidizado. Empleo de un eyector para varios filtros

    Energy Technology Data Exchange (ETDEWEB)

    Sancho Rod, J; Rodirgo Otero, A

    1966-07-01

    Tests to know the efficiency of a porous filters cleaning system by blow-back that uses on ejector for each set of simultaneously cleaned filters were carried out. A Calculation method to obtain the optimum ejector for this system was shown, taking n=2, as optimum number of working for the fluidized bed reactors belonging to the Pilot plant of the Materials Division at JEN. That is two filters for each ejector. (Author)

  7. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  8. TAR-1 A programme for the determination of time behaviour of neutron density on a thermal reactor; TAR-1. Programa para la determinacion del comportamiento temporal de la densidad neutronica en reactor termico

    Energy Technology Data Exchange (ETDEWEB)

    Torres Vida, J

    1963-07-01

    This programme, written for the UNIVAC-UCT of J.E.N., obtain the time behaviour of neutron density as a function of both positive and negative step change in reactivity. These results are obtained from solutions of the space-independent kinetic equations of a bare thermal reactor based on the Fermi continuous slowing down model and using six groups of delayed neutrons. (Author) 3 refs.

  9. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  10. 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

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  13. Clique aqui para Decidir: As Recomendacoes Online Na Decisao de Compra dos Servicos Hoteleiros

    Directory of Open Access Journals (Sweden)

    Raissa Karen Leitinho Sales

    2015-01-01

    Full Text Available Este estudo investigou como as opinioes publicadas em sites de recomendacoes influenciam o processo de decisao de compra de servicos hoteleiros. Na coleta de dados, pelo metodo qualitativo, utilizou‑se entrevistas com internautas e consumidores que ja se sentiram influenciados pelas opinioes de outros usuarios. Os resultados indicam que as opinioes postadas atuam de forma relevante na decisao de compra de servicos hoteleiros. Os principais fatores para tornar os comentarios dos internautas referencias sao a credibilidade e a confianca que os buscadores de informacoes lhes atribuem. Identificaram‑se, ainda, os processo de compartilhamento e de busca de informacoes online dos servicos turisticos.

  14. Simplified methodology for control cell constant calculations of the reactor cores for the space kinetics; Metodologia simplificada para calculos das constantes das celulas de controles dos nucleos de reatores para a cinetica espacial

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Rubens Souza dos [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Martinez, Aquilino Senra; Alvim, Antonio Carlos Marques [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    In this work is presented a methodology which focuses the distribution of neutron absorber rods in nuclear reactor power plants, for utilizing in space kinetic calculations, principally in the cluster ejection transients of control rods. A numerical model for macroscopic constant calculations based on the knowledge of the neutron flux without the control rods is proposed, as alternative to the analytical models, based on the hypothesis of the null current on the cell super boundaries. The proposed model in this work has itself showed adequate to deal with problems with strong space dependence, once that the model showed consistence in the global average built in the analytical model. (author)

  15. Uranium hexafluoride reconversion used for dispersion fuel elements fabrication for IEAR-1/SP reactor; Reconversao de hexafluoreto de uranio para a fabricacao de combustiveis na forma de dispersoes para o reator IEA-R1/SP

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, E.F. Urano de; Lainetti, P.E.; Gomes, R.P. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    1996-07-01

    In this paper are described the main chemical process employed in the Chemical Processes Division of the Fuel Technology Department - IPEN for conversion of enriched UF{sub 6} in ammonium diuranate - DUA and uranium tetrafluoride - UF{sub 4}. These activities have assured the continuity of fuel elements production at IPEN since 1984. The uranium recovery from scraps of the fuel elements production and the purification processes are also described. Those compounds are important intermediate products in the fabrication routine and in development dispersed fuel elements with higher uranium loading for IEA{sub R}1 research reactor power increase program. (author)

  16. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor; Algoritmo neuro-difuso para la identificacion de la potencia neutronica del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.; Benitez R, J. S. [Instituto Tecnologico de Toluca, Division de Estudios de Posgrado e Investigacion, Av. Tecnologico s/n, Ex-Rancho La Virgen, 50140 Metepec, Estado de Mexico (Mexico); Segovia de los Rios, J. A.; Rivero G, T. [ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jorge.benitez@inin.gob.mx

    2009-10-15

    In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

  17. Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core; Programas de captura, analisis y graficado de datos para el estudio de la termometria del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C

    1991-05-15

    This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

  18. Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2014-07-01

    The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

  19. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  20. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Barranco R, F.

    2015-07-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  1. Methodology for solving the equation of transport ordered discrete TORT code in the reactor IPEN/MB-01; Metodologia para resolver la ecuacion del transporte con el codigo de Ordenadas Discretas TORT en el reactor IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, A.; Abarca, A.; Barrachina, T.; Miro, R.; Verdu, G.

    2013-07-01

    The resolution of the neutron transport equation in steady state in pool-type nuclear reactors, is normally achieved through 2 different numerical methods: Monte Carlo (stochastic) and discrete ordinates (deterministic). The discrete ordinates method solves the neutron transport equation for a set of specific addresses, obtaining a set of equations and solutions for each direction, where the solution for each direction is the angular flux. With the aim of treating energy dependence, used energy multigroup approximation, thus obtaining a set of equations that depends on the number of energy groups considered.

  2. Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies; Determination de l'Efficacite des Barres de Reglage dans les Ensembles Combustibles du reacteur VVER; Opredelenie ehffektivnosti reguliruyushchikh sterzhnej v sborkakh reaktora VVEHR; Determinacion de la Eficacia de las Barras de Control en los Conjuntos de Elementos Combustibles del Reactor VVER

    Energy Technology Data Exchange (ETDEWEB)

    Semenov, V. N.; Lunin, G. L.; Komissarov, L. V.; Kamyshan, A. N.; Halizev, V. I.; Andrianov, G. Ja.; Voznesenskij, V. A.; Kuz' micheva, V. A.; Lebedev, V. I. [Ordena Lenina Institut Atomnoj Energii Im. I.V. Kurchatova, Moskva, SSSR (Russian Federation)

    1964-06-15

    The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author) [French] Le memoire decrit les experiences qui ont ete faites pour determiner l 'efficacite des absorbants contenus dans les barres de compensation, l'effet cavitaire et l 'efficacite des absorbants gaines de materiaux divers, au moyen d'assemblages homogenes de cartouches de combustible du reacteur VVER (reacteur de puissance ralenti et refroidi a l 'eau ayant le meme taux d'enrichissement. On y trouve en outre des donnees sur certaines experiences executees a l 'aide d'assemblages de cartouches de combustible taux d'enrichissement differents. Ces travaux permettent de verifier la methode de calcul et d'evaluer ses possibilites d'application aux reacteurs non homogenes. (author) [Spanish] Se describen en la memoria experimentos para determinar la eficacia de los materiales absorbentes contenidos en las barras de compensacion, el efecto de cavitacion y la eficacia de los materiales absorbentes revestidos de diversos materiales, realizados con ayuda de los conjuntos homogeneos de elementos combustibles del reactor VVER (reactor de potencia moderado y refrigerado por agua) con un solo grado de enriquecimiento. Ademas, se exponen datos sobre los experimentos efectuados con ayuda de conjuntos de grados de enriquecimientos; variados. Tales experimentos permiten verificar el metodo de calculo teorico, utilizad o y evaluar la posibilidad de aplicarlo a los reactores no homogeneos. (author

  3. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries; Definicion y Analisis de Benchmarks de Reactores de Agua Pesada para Pruebas de Nuevas Bibliotecas de Datos Wims-D

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, Francisco [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

    2000-07-01

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable.

  4. Comparison of results for burning with BWR reactors CASMO and SCALE 6.2 (TRITON / NEWT); Comparacion de los resultados de quemado para reactores BWR con CASMO y SCALE 6.2 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

    2014-07-01

    In this paper we compare the results from two codes burned, CASMO and SCALE 6.2 (TRITON). To do this, is simulated all segments corresponding to a boiling water reactor (BWR) using both codes. In addition, to account for different working points, simulations changing the instantaneous variables, these are repeated: void fractions (6 points), fuel temperature (6 points) and control rods (two points), with a total of 72 possible combinations of different instantaneous variables for each segment. After all simulations are completed for each segment, we can reorder the obtained cross sections, as SCALE CASMO both, to create a library of compositions nemtab format. This format is accepted by the neutronic code of nodal diffusion, PARCS v2.7. Finally compares the results obtained with PARCS and with the SIMULATE3 -SIMTAB methodology to level of full reactor. Also, we have made use of the KENO-VI and MCDANCOFF modules belonging to SCALE. The first is a Monte Carlo transport code with which you can validate the value of the multiplier, the second has been used to obtain values of Dancoff factor and increase the accuracy of model SCALE. (Author)

  5. Application of COMSOL in the solution of the neutron diffusion equations for fast nuclear reactors in stationary state; Aplicacion de COMSOL en la solucion de las ecuaciones de difusion de neutrones para reactores nucleares rapidos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Silva A, L.; Del Valle G, E., E-mail: evalle@ipn.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico)

    2012-10-15

    This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S{sub 2} approximation. (Author)

  6. Subsurface microbial ecology. Epi fluorescence direct counts; Ecologia microbica del sottosuolo: metodo di conta diretta in epifluorescenza

    Energy Technology Data Exchange (ETDEWEB)

    Barra Caracciolo, A.; Silvestri, C.; Creo, C.; Izzo, G. [ENEA, Centro Ricerche Casaccia, Rome (Italy). Dipt. Ambiente

    1998-07-01

    To the aim of recognize the importance of microorganisms in affecting or even determining the fate of xenobiotics in the subsurface environment evaluating bacteria concentration in a subsurface ecosystem, the report discusses a soil sample treatment method which has been developed for epi fluorescence direct counting with DAPI. [Italian] Lo studio discute un metodo di trattamento del campione per la conta diretta in epifluorescenza con un marcatore selettivo per il DNA, il DAPI, al fine di quantificare la concentrazione batterica del sottosuolo e studiare il ruolo dei microrganismi nella biodegradazione delle molecole esogene, ancora poco indagato.

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

    OpenAIRE

    Andrea Pérez; Patricia Torres

    2008-01-01

    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 presenc...

  8. Modelos matemáticos para reactores biológicos de lecho empacado (pbr): una revisión bibliográfica

    OpenAIRE

    Corredor, Deisy; Caicedo, Luis Alfonso

    2010-01-01

    El modelo matemático y análisis teórico de reactores biológicos de lecho empacado (PBR) ha sido estudiado por diferentes autores, quienes tuvieron en cuenta variedad de cinéticas de reacción, modelos unidimensionales, homogéneos, pseudohomogéneos y heterogéneos. Las ecuaciones resultantes del modelo fueron solucionadas, en su gran mayorIa, por sistemas de métodos numéricos. Se ha analizado en estos el efecto de variables de proceso de importancia fIsica con respecto a parámetros de diseno y o...

  9. Fuel management inside the reactor. Report of generation of the nuclear bank for the fuel of the initial load of the Laguna Verde U-1 reactor with the FMS codes; Administracion de combustible dentro del reactor. Reporte de generacion del banco nuclear para el combustible de la carga inicial del reactor de Laguna Verde U-1 con los codigos del FMS

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Torres A, C. [CFE, Veracruz (Mexico)

    1991-06-15

    In this work in a general way the form in that it was generated the database of the initial fuel load of the Laguna Verde Unit 1 reactor is described. The initial load is formed with fuel of the GE6 type. The obtained results during the formation of the database in as much as to the behavior of the different cell parameters regarding the one burnt of the fuel and the variation of vacuums in the coolant channel its are compared very favorably with those reported by the General Electric fuel supplier and reported in the design documents of the same one. (Author)

  10. Preparation of hydrotalcite compounds using ultrasound irradiation to capture CO{sub 2}; Preparacion de compuestos tipo hidrotalcita utilizando irradiacion de ultrasonido para la captura de CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.A.J.; Paredes, S.P.; Valenzuela, M.A.; Hernandez, M.L. [Instituto Politecnico Nacional, ESIQIE, Mexico, D.F. (Mexico)]. E-mail: sparedesc@ipn.com.mx

    2009-09-15

    Al-Mg hydrotalcite compounds (HTC) were prepared using co-precipitation, sol-gel and reconstruction of the structure with ultrasound-assisted irradiation. The interlaminar components for each method were nitrate, acetylacetonate ethoxide and metavanadate, respectively. Optimization of the synthesis was performed using x-ray diffraction. The effect of the different parameters on synthesis was studied, including pH, time and ultrasound irradiation power. In addition, for the reconstruction method, temperature and calcination time were evaluated. For all methods, ultrasound-assisted methods were found to be more efficient and economical than conventional methods reported (autoclave). They also have the advantage of being able to control properties such as crystallinity, porosity and the specific surface, which significantly depends on the preparation method, irradiation time and type of interlaminar component. These methods are intended to synthesize interlaminar anionic materials that are very scarce in nature with better properties than traditional adsorbents used for the capture of CO{sub 2}. [Spanish] Se prepararon compuestos tipo hidrotalcita Al-Mg por los metodos de: coprecipitacion, sol gel y reconstruccion de la estructura asistidos por irradiacion de ultrasonido. Los componentes interlaminares para cada metodo fueron respectivamente: nitrato, etoxido-acetilacetonato y metavanadato. La optimizacion de la sintesis, se efectuo mediante difraccion de rayos-X. Se estudio el efecto de diversos parametros en la sintesis: pH, tiempo y potencia de irradiacion de ultrasonido, ademas, para el metodo de reconstruccion se evaluaron la temperatura y el tiempo de calcinacion. En todos los casos se encontro que el empleo de metodos asistidos por ultrasonido resultan ser mas eficientes y economicos que los metodos convencionales reportados (autoclave), ademas tienen la ventaja, de poder controlar propiedades tales como: la cristalinidad, la porosidad y la superficie

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  12. 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...

  13. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor; Desarrollo de un software para el control del sistema de gestion de calidad del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Herrera A, E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Hernandez, L.V.; Hernandez, J.A. [UAEM, Depto. de Ingenieria en Computacion, 50000 Toluca, estado de Mexico (Mexico)]. e-mail: eha@nuclear.inin.mx

    2006-07-01

    The quality has not only become one of the essential requirements of the product but rather at the presenme it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  14. Modeling of an immobilized lipase tubular reactor for the production of glycerol and fatty acids from oils; Modelado de un reactor tubular de lipasas inmovilizadas para la produccion de glicerol y acidos grasos a partir de aceites

    Energy Technology Data Exchange (ETDEWEB)

    Oddone, S.; Grasselli, M.; Cuellas, A.

    2010-07-01

    Advances in the design of a bioreactor in the fats and oils industry have permitted the hydrolysis of triglycerides in mild conditions and improved productivity while avoiding the formation of unwanted byproducts. The present work develops a mathematical model that describes the hydrolytic activity of a tubular reactor with immobilized lipases for the production of glycerol and fatty acids from the oil trade. Runge Kuttas numerical method of high order has been applied, considering that there is no accumulation of the substratum in the surface of the membrane, where the enzyme is. At the same time, different equations based on the kinetic model of Michaelis Mentens and the Ping-Pong bi-bi mechanism were examined. Experimental data in discontinuous systems are the basis for the development of the quantitative mathematical model that was used to simulate the process computationally. The obtained results allow for optimizing both the operative variables and the economic aspects of industrial processes. (Author)

  15. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ; Diseno asistido por computadora (DAC) para mejorar la electronica de los canales nucleares del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jlgm@nuclear.inin.mx

    2007-07-01

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  16. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H; Ortiz V, J [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  17. Evaluación de los Parametros Fisicoquímicos en Reactores Discontinuos de Lodos Activados para el Tratamiento de Aguas con Metanol.

    Directory of Open Access Journals (Sweden)

    Marvin Carabali Lasso

    2017-01-01

    Full Text Available Contexto: en la actualidad existen varios tipos de tecnologías limpias que ayudan a la descontaminación del agua residual industrial, como los sistemas de lodos activados. Dichas tecnologías han resultado efectivas en la remoción de materia orgánica y coloidal en aguas residuales domésticas e industriales, biosorción de metales pesados y remoción de patógenos y nutrientes (N y P. Lo anterior motiva la propuesta de trabajar con un sistema de tratamiento de lodos activados a escala en un ambiente controlado, a fin de demostrar su efectividad en tratamiento de aguas con metanol. Método: el sistema de tratamiento de lodos activados se diseña y construye a escala a nivel de laboratorio para estudiar diferentes concentraciones de metanol en aguas residuales de una industria. En los montajes de cinco sistemas de tratamiento se realizan mediciones de parámetros fisicoquímicos como: pH, oxígeno disuelto, temperatura, demanda química de oxígeno, sólidos suspendidos totales y la pérdida de metanol; a las 0, 24 y 48 h por cuadruplicado para observar el comportamiento de dichos parámetros durante la reducción de metanol. Resultados: para los sistemas de lodos activados se demuestra una pérdida promedio de metanol del 10,3 %. Los promedios de oxígeno disuelto decrecieron en el tiempo, aunque la demanda química de oxígeno es variable. Cada sistema con concentración diferente de metanol presenta comportamiento particular. Conclusiones: esta investigación indica que la biodegradación de metanol en aguas residuales industriales es posible, teniendo en cuenta que en el sistema de tratamiento a escala se obtuvo una disminución de metanol en el tiempo. Además, los resultados motivan a continuar investigando sobre esta tecnología limpia, de la cual no se dispone suficiente literatura en lo que respecta al manejo de lodos activados para la eliminación de metanol.

  18. Trends in the analytical methods for water analysis; Tendencias de los metodos analiticos para el analisis del agua

    Energy Technology Data Exchange (ETDEWEB)

    Tortajada-Genero, L. A.; Campins-Falco, P. [Universidad de Valencia (Spain)

    2003-07-01

    The technological and methodological advances have provided and evolution of analytical chemistry, specially in the area of water analysis. There area wide variety of problems for this kind of analysis, as result of the high number of possible objectives. They involve from complex environmental studies to routine control of a physical-chemical parameter. Additionally, the variety of analysis to analyse is increasing as in total number as in complexity terms. The response has been the appearance of a high number of analytical methods. Some guidelines and recommendations have been described together with different strategies for quality assurance. In this paper, a review of the last analytical methods described in the scientific references for determinations in water was realised. The evolution and future perspectives were also studied. (Author) 16 refs.

  19. Experimental method for calculation of effective doses in interventional radiology; Metodo experimental para calculo de dosis efectivas en radiologia intervencionista

    Energy Technology Data Exchange (ETDEWEB)

    Herraiz Lblanca, M. D.; Diaz Romero, F.; Casares Magaz, O.; Garrido Breton, C.; Catalan Acosta, A.; Hernandez Armas, J.

    2013-07-01

    This paper proposes a method that allows you to calculate the effective dose in any interventional radiology procedure using an anthropomorphic mannequin Alderson RANDO and dosimeters TLD 100 chip. This method has been applied to an angio Radiology procedure: the biliary drainage. The objectives that have been proposed are: to) put together a method that, on an experimental basis, allows to know dosis en organs to calculate effective dose in complex procedures and b) apply the method to the calculation of the effective dose of biliary drainage. (Author)

  20. Tools for building Breast Cancer CAD methods; Herramientas para elaborar metodos de CAD del cancer de mama

    Energy Technology Data Exchange (ETDEWEB)

    Diaz-Herrero, G.; Franco-Valiente, J. M.; Suarez-Ortega, C.; Rubio del Solar, M.; Ramos-Pollam, R.; Guevara-Lopez, M. A.; Gonzalez de Posada, N.; Ramos, I.; Loureiro, J.

    2011-07-01

    This paper describes the main results of the collaboration currently ongoing between CETA-CIEMAT and the Faculty of Engineering at the University of Porto. It introduces the Mammography image Workstation for Analysis and Diagnosis (MIWAD) and the data analysis processes through which Grid infrastructures are used to develop Breast Cancer Computer-Aided Diagnosis (CAD) methods.MIWAD is a novel integrated software framework that integrates a specialized graphical user interface combining digital image processing, pattern recognition and artificial intelligence techniques. This paper describes its successful application in a pilot experience, on the first Portuguese Breast Cancer Digital Repository. (Author)

  1. A proactive method for safety management in nuclear facilities; Um metodo proativo para gerenciamento da seguranca em instalacoes nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Grecco, Claudio Henrique dos Santos; Carvalho, Paulo Victor Rodrigues de; Santos, Isaac Antonio Luquetti dos, E-mail: grecco@ien.gov.br, E-mail: paulov@ien.gov.br, E-mail: luquetti@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN/RJ), Rio de Janeiro, RJ (Brazil). Div. de Instrumentacao e Confiabilidade Humana

    2014-07-01

    Due to the modern approach to address the safety of nuclear facilities which highlights that these organizations must be able to assess and proactively manage their activities becomes increasingly important the need for instruments to evaluate working conditions. In this context, this work presents a proactive method of managing organizational safety, which has three innovative features: 1) the use of predictive indicators that provide current information on the performance of activities, allowing preventive actions and not just reactive in safety management, different from safety indicators traditionally used (reactive indicators) that are obtained after the occurrence of undesired events; 2) the adoption of resilience engineering approach in the development of indicators - indicators are based on six principles of resilience engineering: top management commitment, learning, flexibility, awareness, culture of justice and preparation for the problems; 3) the adoption of the concepts and properties of fuzzy set theory to deal with subjectivity and consistency of human trials in the evaluation of the indicators. The fuzzy theory is used primarily to map qualitative models of decision-making, and inaccurate representation methods. The results of this study aim an improvement in performance and safety in organizations. The method was applied in a radiopharmaceutical shipping sector of a nuclear facility. The results showed that the method is a good monitoring tool objectively and proactively of the working conditions of an organizational domain.

  2. Sintese fotoquimica de polipirrol em membranas microporosas de PVDF : um novo metodo para obter compositos condutores eletricos

    OpenAIRE

    Ruth Marlene Campomanes Santana

    1995-01-01

    Resumo: Este trabalho descreve em linhas gerais um novo método de obter compósitos condutores de polipirrol (PPy)/Poli(fluoreto de vinilideno) (PVDF), polimerização via luz UV induzida em presença de um adequado fotogerador de ácido (PAG), compostos tais como os sais de Ferro-areno e Trifenil sulfônio ou compostos geradores orgânicos tal como a triazina clorinada. Estudos Gravimétricos confirmados por Análise Termogravimé-trico (TGA), tem mostrado que o rendimento da polimerização insidente n...

  3. Manual method for dose calculation in gynecologic brachytherapy; Metodo manual para o calculo de doses em braquiterapia ginecologica

    Energy Technology Data Exchange (ETDEWEB)

    Vianello, Elizabeth A.; Almeida, Carlos E. de [Instituto Nacional do Cancer, Rio de Janeiro, RJ (Brazil); Biaggio, Maria F. de [Universidade do Estado, Rio de Janeiro, RJ (Brazil)

    1998-09-01

    This paper describes a manual method for dose calculation in brachytherapy of gynecological tumors, which allows the calculation of the doses at any plane or point of clinical interest. This method uses basic principles of vectorial algebra and the simulating orthogonal films taken from the patient with the applicators and dummy sources in place. The results obtained with method were compared with the values calculated with the values calculated with the treatment planning system model Theraplan and the agreement was better than 5% in most cases. The critical points associated with the final accuracy of the proposed method is related to the quality of the image and the appropriate selection of the magnification factors. This method is strongly recommended to the radiation oncology centers where are no treatment planning systems available and the dose calculations are manually done. (author) 10 refs., 5 figs.

  4. Implementation of statistical analysis methods for medical physics data; Implementacao de metodos de analise estatistica para dados de fisica medica

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Marilia S.; Pinto, Nivia G.P.; Barroso, Regina C.; Oliveira, Luis F., E-mail: mariliasilvat@gmail.co, E-mail: lfolive@oi.com.b, E-mail: cely_barroso@hotmail.co, E-mail: nitatag@gmail.co [Universidade do Estado do Rio de Janeiro (UERJ), Rio de Janeiro, RJ (Brazil). Inst. de Fisica

    2009-07-01

    The objective of biomedical research with different radiation natures is to contribute for the understanding of the basic physics and biochemistry of the biological systems, the disease diagnostic and the development of the therapeutic techniques. The main benefits are: the cure of tumors through the therapy, the anticipated detection of diseases through the diagnostic, the using as prophylactic mean for blood transfusion, etc. Therefore, for the better understanding of the biological interactions occurring after exposure to radiation, it is necessary for the optimization of therapeutic procedures and strategies for reduction of radioinduced effects. The group pf applied physics of the Physics Institute of UERJ have been working in the characterization of biological samples (human tissues, teeth, saliva, soil, plants, sediments, air, water, organic matrixes, ceramics, fossil material, among others) using X-rays diffraction and X-ray fluorescence. The application of these techniques for measurement, analysis and interpretation of the biological tissues characteristics are experimenting considerable interest in the Medical and Environmental Physics. All quantitative data analysis must be initiated with descriptive statistic calculation (means and standard deviations) in order to obtain a previous notion on what the analysis will reveal. It is well known que o high values of standard deviation found in experimental measurements of biologicals samples can be attributed to biological factors, due to the specific characteristics of each individual (age, gender, environment, alimentary habits, etc). This work has the main objective the development of a program for the use of specific statistic methods for the optimization of experimental data an analysis. The specialized programs for this analysis are proprietary, another objective of this work is the implementation of a code which is free and can be shared by the other research groups. As the program developed since the method modeling through his implementation, this fact will allows that modifications and adaptations be accomplished any time at null cost

  5. Methods to identify and locate spent radiation sources; Metodos para la identificacion y localizacion de fuentes radiactivas gastadas

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    The objective of this manual is to provide essential guidance to Member States with nuclear applications involving the use of a wide range of sealed radiation sources on the practical task of physically locating spent radiation sources not properly accounted for. Advice is also provided to render the located source safe on location. Refs, figs, tabs.

  6. Methods of calculus for neutron spectrometry in proportional counters; Metodos de calculo para espectrometria de neutrones en contadores proporcionales

    Energy Technology Data Exchange (ETDEWEB)

    Butragueno, J L; Blazquez, J B; Barrado, J M

    1976-07-01

    Response functions for cylindrical proportional counters with hydrogenated gases have been determined, taking in account only wall effect, by means of two independent calculus methods. One of them is a Monte Carlo application and the other one analytica at all. Results of both methods have been compared. (Author)

  7. The Brazilian dilution method for ballast water exchange; O metodo de diluicao brasileiro para troca de agua de lastro

    Energy Technology Data Exchange (ETDEWEB)

    Mauro, Celso Alleluia [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas. Avaliacao e Monitoramento Ambiental]. E-mail: celso@cenpes.petrobras.com.br; Land, Claudio Goncalves [PETROBRAS, Rio de Janeiro, RJ (Brazil). Abastecimento, Logistica e Planejamento]. E-mail: cgland@petrobras.com.br; Pimenta, Jose Maria Hollanda Alvares; Barreto, Francisco Carlos Peixoto [PETROBRAS, Rio de Janeiro, RJ (Brazil). Engenharia; Brandao, Marcus Vinicius Lisboa; Marroig, Nilton Lemos [Transpetro, Rio de Janeiro, RJ (Brazil). Frota Nacional de Petroleiros. Inspetoria Geral; Tristao, Maria Luiza Braganca [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas. Quimica; Fadel, Andre Luiz da Fonseca [PETROBRAS, Rio de Janeiro, RJ (Brazil). Financas Corporativa e Tesouraria; Villac, Maria Celia; Fernandes, Lohengrin; Paranhos, Rodolfo; Dias, Cristina; Bonecker, Sergio; Denise Tenenbaum [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Biologia; Persich, Graziela; Garcia, Virginia; Odebrecht, Clarisse [Fundacao Universidade do Rio Grande, RS (Brazil). Dept. de Oceanografia

    2002-12-01

    In a precautionary approach and dealing with the coming International Maritime Organization (IMO) regulations on ballast water, PETROBRAS developed a new method for ballast water exchange in tankers. Differently from ordinary methods PETROBRAS method which have been called Brazilian Dilution Method (BDM) or Dilution Method involves ballast loading through the top with simultaneous unloading from the bottom of the tanks. The method proposal was firstly presented to IMO, which encouraged PETROBRAS to carry out a field trial. PETROBRAS in June 1998 carried out a trial in the product carrier M/V Lavras. A simulation study was useful to plan the trial assessing the theoretical efficiency of the method, establishing the best sampling points and comparing the BDM with the Tank Overfilling Method (TOM). Simulation showed that for the same tank shape, the water renewal in BDM is more effective than in TOM and that 90 % of water renewal could be obtained by BDM. A dye concentration variation monitoring and a biological assessment were performed and the results confirmed that over than 90 % of the ballast water was renewed after three exchanges. The method was proved safe, practical, economical and suitable to minimize the risk of exotic species transport between ports. (author)

  8. A definitive method for dosimetry with radiochromic movie; Un metodo definitivo para la dosimetria con pelicula radiocromica

    Energy Technology Data Exchange (ETDEWEB)

    Miras del Rio, H.; Arrans Lara, R.

    2013-07-01

    Despite the undoubted benefits of radiochromic film as a two-dimensional radiation detector, the manufacturing process and characteristics of scanners make to present some inherent difficulties that result that its use is complex, tedious, and if not performed so appropriate, imprecise. A protocol that simplifies the calibration process while reducing the number of films used for this purpose is proposed systematically corrects several difficulties, both inherent in the manufacture of the film as the reader and, what is more important achieves levels comparable to those of other more widespread use detectors accuracy. (Author)

  9. Thermochemical methods for the treatment of oil contaminated sand; Metodo termoquimico para tratamento de areia contaminada por oleo

    Energy Technology Data Exchange (ETDEWEB)

    Pimenta, Rosana C.G.M. [Fundacao Jose Bonifacio, Rio de Janeiro, RJ (Brazil); Khalil, Carlos N. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

    2003-07-01

    The Nitrogen Generating System (SGN in Portuguese) is a thermochemical method first developed for cleaning and removal of paraffin deposits in production and export pipelines. SGN is based on a redox chemical reaction between two salts which is catalyzed in acidic pH. The reaction is strongly exothermic and its products are nitrogen, sodium chloride, water and heat. All reaction products are harmless to the environment. In January 2000 there was a major oil spill in Guanabara Bay, Rio de Janeiro, which contaminated 2400 tons of sand. This work, developed at PETROBRAS Research Center (CENPES), was based on SGN technology which has been adapted for cleaning contaminated sand and recovering of spilled oil. By combining simultaneous effects of the SGN treatment such as heating, turbulence and floatation, one can remove, within 98% of efficiency, spilling oil from contaminated sand and removed oil can be securely returned to refining process. SGN technology has proved to be efficient, fast, low cost and ecologically correct method for cleaning contaminated sand and can be applied in loco right after a contamination event. (author)

  10. Thermochemical method for the treatment of oil contaminated sand; Metodo termoquimico para tratamento de areia contaminada por oleo

    Energy Technology Data Exchange (ETDEWEB)

    Pimenta, Rosana C.G.M. [Fundacao Gorceix, Ouro Preto, MG (Brazil)]|[PETROBRAS S.A., Rio de Janeiro, RJ (Brazil); Khalil, Carlos N. [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas (CENPES)

    2004-07-01

    In January 2000 there was a major oil spill in Guanabara Bay, Rio de Janeiro, which contaminated 2400 tons of sand. This work, based on NGS (Nitrogen Generating System) technology, was adapted for cleaning contaminated sand and recovering of spilled oil. NGS is a thermochemical method first developed for removal of paraffin deposits in production and export pipelines. The method is based on a strongly exothermic redox chemical reaction between two salts catalyzed in acidic pH. The reaction products are harmless to the environment and consist of nitrogen, sodium chloride, water and heat. By combining simultaneous effects of the treatment such as heating, turbulence and floatation, one can remove, within 98% of efficiency, spilling oil from contaminated sand. After treatment, removed oil can be securely returned to refining process. The method has proved to be efficient, fast, low cost and ecologically correct method for cleaning contaminated sand and can be applied in place right after a contamination event. (author)

  11. A method multi criterio to evaluate projects of rural electrification; Un metodo multicriterio para evaluar proyectos de electrificacion rural

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez Posse, E

    1994-07-01

    In this document about the problem of the evaluation projects methodologies in rural electrification.The low analysis problem is of complex nature, because each project is evaluation object and an economic agent. One of these agents identifies different benefits and cost, and also has a different approaches for value them.In consequence, the form in that it is carried out the evaluation of the one project for each one of this agents that it is usually solved for mechanisms linked to the capacity of incidence or of determination of each one of them, this does not assures a satisfactory results for the general interest.

  12. A new method for dosimetry with films radiochromic; Un nuevo metodo para la dosimetria con peliculas radiocromica

    Energy Technology Data Exchange (ETDEWEB)

    Mendez Carot, I.

    2013-07-01

    in this paper a new method is presented and the results of the comparison between the calibration is summarized based on a planning reference and calibration obtained from the irradiated fragments measure different dose levels multichannel compare dosimetry based on the weighted average dosimetry described by Micke et al.(present in the FilmQAPro software) and, finally, show different results obtained with the method proposed in several applications clinics. (Author)

  13. Determination of power distribution in reactor with nodal expansion method; Izrachun porazdelitve mochi v reaktorju z metodo nodalne ekspanzije

    Energy Technology Data Exchange (ETDEWEB)

    Kromar, M; Trkov, A [Institut Jozef Stefan, Ljubljana (Yugoslavia); Pregl, G [Tehnishka Fakulteta Maribor Univ. (Yugoslavia)

    1988-07-01

    Nodal expansion method (NEM) is one of the advanced coarse-mesh methods based on integral form of few-group diffusion equation. NEM can be characterized by high accuracy and computational efficiency. Method was tested by development of computer code NEXT. Validation of the code was performed by calculation of 2-D and 3-D IAEA benchmark problem. NEXT was compared with codes based on other methods (finite differences, finite elements) and has been found to be accurate as well as fast. (author)

  14. The Economical Application of Non-Destructive Testing to Reactor Components, Especially Jacket Tubing; Avantages Economiques du Controle Non Destructif des Pieces de Reacteurs, Notamment des Tubes de Gainage; Ehkonomicheskoe primenenie nedestruktivnykh ispytanij dlya reaktornykh komponentov, v chastnosti obolochechnykh trub; Aplicacion en Condiciones Economicas de Ensayos No Destructivos a las Piezas de los Reactores, en Especial a los Tubos de Revestimiento

    Energy Technology Data Exchange (ETDEWEB)

    Renken, C. J. [Metallurgy Division Argonne National Laboratory Argonne, IL (United States)

    1965-10-15

    . Des indications erronees de defauts contribuent directement a l'accroissement du prix de revient des pieces; c'est pourquoi le memoire contient une evaluation de ces effets pour les methodes ultrasonores et electromagnetiques en ce qui concerne plusieurs sources frequentes d'indications erronees. L'auteur expose l'experience acquise au Laboratoire national d'Argonne dans l'application de ces methodes a des quantites relativement importantes de tubes d'origines diverses, du point de vue du prix minimum du controle parunite de longueur de tube. Cette partie du memoire resume egalement l'experience acquise au Laboratoire d'Argonne avec les methodes electromagnetiques et impulsions les plus recentes. L'auteur discute l'influence primordiale, mais generalement trop negligee, du diametre et de l'epaisseur du tube sur le prix de revient du controle. Comme la question de l'economie du controle est etroitement liee et celle des defauts admissibles, l'auteur expose les normes appliquees a cet egard au Laboratoire d'Argonne. Enfin, il enumere les obstacles pratiques et theoriques qui empechent de reduire le prix de revient du controle des pieces et il s'efforce de faire une prevision des reductions possibles de c e prix grace aux methodes ultiasonores et electromagnetiques. (author) [Spanish] Ademas de las caracteristicas que debe reunir el modelo ideal de reactor, hay que aplicarle metodos de ensayo que no tengan caracter destructivo. Como otros ideales, es probable que este no se alcance nunca. Para cualquier modelo en el que el costo sea un factor importante, la cuestion de la posibilidad de ensayar las piezas en condiciones economicas debe plantearse al mismo tiempo que la de la posibilidad de fabricacion. En la presente memoria se resellan algunas observaciones al respecto y se examina la importancia que ha de atribuirse a los metodos de ensayo no destructivo al establecer las especificaciones correspondientes. El fabricante ademas es responsable de la utilizacion de

  15. Sintering study in vertical fixed bed reactor for synthetic aggregate production; Estudo da sinterizacao em reator vertical de leito fixo para producao de agregado sintetico

    Energy Technology Data Exchange (ETDEWEB)

    Quaresma, D.S.; Neves, A.S.S.; Melo, A.O.; Pereira, L.F.S.; Bezerra, P.T.S.; Macedo, E.N.; Souza, J.A.S., E-mail: danysq@gmail.com [Universidade Federal do Para (UFPA), Belem, PA (Brazil). Faculdade de Engenharia Quimica

    2017-04-15

    The synthetic aggregates are being employed in civil construction for the reduction of mineral extraction activities. Within this context, the recycling of industrial waste is the basis of the majority of processes to reduce the exploitation of mineral resources. In this work the sintering in a vertical fixed bed reactor for synthetic aggregate production using 20% pellets and 80% charcoal was studied. The pellets were prepared from a mixture containing clay, charcoal and fly ash. Two experiments varying the speed of air sucking were carried out. The material produced was analyzed by X-ray diffraction, scanning electron microscopy, measures of their ceramic properties, and particle size analysis. The results showed that the solid-state reactions, during the sintering process, were efficient and the produced material was classified as coarse lightweight aggregate. The process is interesting for the sintering of aggregates, and can be controlled by composition, particle size, temperature gradient and gaseous flow. (author)

  16. 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

  17. Dispersions of Oxides in Oxide Matrices as High-Temperature Reactor Fuels; Dispersions d'oxyde dans des matrices d'oxyde, utilisees comme combustibles dans des reacteurs a haute temperature; Dispersiya okisej v okislovykh matritsakh v kachestve topliva dlya vysokotemperaturnogo reaktora; Empleo de dispersiones de oxidos en matrices de oxidos, como combustibles para reactores de elevada temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Williams, J. [Atomic Energy Research Establishment, Harwell (United Kingdom)

    1963-11-15

    posibilidad de utilizar dispersiones de PuO{sub 2}, UO2 y ThO{sub 2} en matrices de BeO, Al{sub 2}O{sub 3}, MgO y SIO{sub 2}, desde el punto de vista de la integridad y elaboracion del combustible. Las dos caracteristicas mas importantes en lo que atane a la integridad del combustible son su estabilidad dimensional y su capacidad de retener los productos de fision. La compatibilidad de los componentes del combustible entre si, y entre ellos y el refrigerante, influye en la estabilidad dimensional, pero en este aspecto, las propiedades de los combustibles en forma de oxido son satisfactorias. A la alteracion de las dimensiones del combustible por irradiacion contribuyen los factores siguientes: deterioracion de la matriz por los neutrones y los fragmentos de fision, deteriorizacion de la fase fisionable/fertil por las radiaciones y gases de los productos de fision acumulados. Las tensiones termicas pueden tambien provocar deformaciones. Los conocimientos que se poseen sobre los mecanismos de atenuacion de tensiones son, sin embargo, insuficientes para permitir un estudio teorico razonable del comportamiento de los materiales. Los estudios sobre la liberacion de productos de fision por los oxidos fisionables/fertiles, realizados tanto en condiciones de irradiacion poco intensa como de elevado grado de combustion, han tratado principalmente de los productos gaseosos, en particular del xenon. Los datos que se poseen sobre el desprendimiento de otros productos de fision son muy escasos, lo mismo que los conocimientos sobre el movimiento de los productos de fision en general a traves de los materiales que podrian utilizarse como matrices. De los estudios realizados sobre la permeabilidad de los oxidos puros sinterizados, se deduce que se deberan alcanzar densidades teoricas de 95%, como minimo, o quiza incluso de 98%, paca eliminar la porosidad de dichas matrices. Se han elaborado una serie de procedimientos para preparar las particulas fisionables/fertiles, para revestirlas y para

  18. Fuel design with low peak of local power for BWR reactors with increased nominal power; Diseno de un combustible con bajo pico de potencia local para reactores BWR con potencia nominal aumentada

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia C, R.; Montes, J.L.; Hernandez, J.L.; Ortiz, J.J.; Castillo, A. [ININ, 52750 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2006-07-01

    The Federal Commission of Electricity recently announcement the beginning of the works related with the increase of the power to 120% of the original nominal one in the Boiling Water Reactors (BWR) of the Laguna Verde Central (CLV): In the National Institute of Nuclear Research (ININ) are carried out studies of the impact on the design of the recharge of derived fuel of this increase. One of the main effects of the power increase type that it is promoting, is the increment of the flow of generated vapor, what takes, to a bigger fraction of vacuum in the core presenting increased values of the maximum fraction to the limit, so much of the ratio of lineal heat generation (XFLPD) as of the ratio of critic power (MFLCPR). In the made studies, it is found that these fractions rise lineally with the increase of the nominal power. Considering that the reactors of the CLV at the moment operate to 105% of the original nominal power, it would imply an increment of the order of 13.35% in the XFLPD and in the MFLCPR operating to a nominal power of 120% of the original one. This would propitiate bigger problems to design appropriately the fuel cycle and the necessity, almost unavoidable, of to resort to a fuel assembly type more advanced for the recharges of the cores. As option, in the ININ the feasibility of continuing using the same type of it fuel assembles that one has come using recently in the CLV, the type GE12 is analyzed. To achieve it was outlined to diminish the peak factor of local power (LPPF) of the power cells that compose the fuel recharge in 13.35%. It was started of a fuel design previously used in the recharge of the unit 1 cycle 12 and it was re-design to use it in the recharge design of the cycle 13 of the unit 1, considering an increase to 120% of the original power and the same requirements of cycle extension. For the re-design of the fuel assembly cell it was used the PreDiCeldas computer program developed in the ININ. It was able to diminish the LPPF

  19. 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

    Radiodosimetria de la Autoridad Regulatoria Nuclear (ARN) utiliza la columna rapida del reactor para el disenio, calibracion y puesta a punto de dosimetros de criticidad asi como de los metodos de evaluacion rapida de las dosis absorbida por el personal en el caso de accidentes. Con este objeto se irradiaron hojuelas de Au(1), Au bajo Cd e In(2), para estimar los flujos neutronicos absolutos termico, epitermico y rapido respectivamente en la posicion de irradiacion. Esta estimacion es tanto mas precisa cuanto mejor se conocen las respuestas de los distintos materiales de los detectores al espectro de neutrones presente. Esto a su vez requiere el conocimiento previo de dicho espectro (dependencia detallada del flujo neutronico con la energia) en la posicion que se analiza. En este trabajo se presenta el calculo neutronico que fue oportunamente requerido al Grupo de Reactores de Investigacion y Conjuntos Criticos de la ARN, con el objeto de determinar el espectro neutronico en la posicion de irradiacion rapida del RA-1. Se ha evaluado el espectro neutronico en diferentes posiciones del reactor, utilizando dos lineas de calculo diferentes y considerando, a los fines de este analisis, una potencia de 40 kW. Se ha representado al reactor y al recinto que lo aloja con un modelo simplificado, unidimensional, como un conjunto de regiones circulares concentricas. Se muestran los graficos de los flujos rapido y termico (con corte en 0,4 eV) en funcion de la distancia al centro del reactor. Se muestra asimismo el grafico del flujo neutronico (en n/cm{sup 2}.seg.eV) en funcion de la energia en la posicion de irradiacion rapida. Se consignan tambien los valores del flujo (en n/cm{sup 2}.seg.eV) en funcion de la energia en otras posiciones tipicas, asi como los valores equivalentes de los flujos integrados (en n/cm{sup 2}.seg). ((1) De acuerdo con la reaccion Au{sup 197}(n, {gamma})Au{sup 198}, con una seccion eficaz de {sigma}{sub 0}=98.8b para neutrones termicos. (2) De acuerdo con la

  20. Advances in the development of the Mexican platform for analysis and design of nuclear reactors: AZTLAN Platform; Avances en el desarrollo de la plataforma mexicana para analisis y diseno de reactores nucleares: AZTLAN Platform

    Energy Technology Data Exchange (ETDEWEB)

    Gomez T, A. M.; Puente E, F. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico); Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: armando.gomez@inin.gob.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2017-09-15

    The AZTLAN platform project: development of a Mexican platform for the analysis and design of nuclear reactors, financed by the SENER-CONACYT Energy Sustain ability Fund, was approved in early 2014 and formally began at the end of that year. It is a national project led by the Instituto Nacional de Investigaciones Nucleares (ININ) and with the collaboration of Instituto Politecnico Nacional (IPN), the Universidad Autonoma Metropolitana (UAM) and Universidad Nacional Autonoma de Mexico (UNAM) as part of the development team and with the participation of the Laguna Verde Nuclear Power Plant, the National Commission of Nuclear Safety and Safeguards, the Ministry of Energy and the Karlsruhe Institute of Technology (Kit, Germany) as part of the user group. The general objective of the project is to modernize, improve and integrate the neutronic, thermo-hydraulic and thermo-mechanical codes, developed in Mexican institutions, in an integrated platform, developed and maintained by Mexican experts for the benefit of Mexican institutions. Two years into the process, important steps have been taken that have consolidated the platform. The main results of these first two years have been presented in different national and international forums. In this congress, some of the most recent results that have been implemented in the platform codes are shown in more detail. The current status of the platform from a more executive view point is summarized in this paper. (Author)

  1. The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR; La importancia de realizar estudios sobre el uso de recombinadores autocataliticos pasivos para control de hidrogeno en reactores tipo ESBWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: jersonsanchez@gmail.com

    2009-10-15

    A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

  2. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors; Una metodologia practica de proteccion radiologica para la reduccion de particulas calientes en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez G, G [Comision Federal de Electricidad, Gerencia del Proyecto Nucleoelectrico Laguna Verde, Disciplina de Fisica Aplicada (Mexico)

    1991-07-01

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  4. Computer program for the calculation of stresses in rotary equipment discs; Programas de computo para el calculo de esfuerzos en discos de equipo rotatorio

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez Delgado, Wilson; Kubiak, Janusz; Serrano Romero, Luis Enrique [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1990-12-31

    In the preliminary design and diagnosis of rotary machines is very common to utilize simple calculation methods for the mechanical and thermal stresses, dynamic and thermodynamic analysis and flow of fluids in this machines (Gutierrez et al., 1989). The analysis with these methods provides the necessary results for the project initial stage of the machine. Later on, more complex tools are employed to refine the design of some machine components. In the Gutierrez report et al., (1989) 34 programs were developed for the preliminary design and diagnosis of rotating equipment; in this article, one of them is presented in which a method for the analysis of mechanical and thermal stresses is applied in discs of uniform or variable thickness that are normally found in turbomachines and rotary equipment. [Espanol] En el diseno preliminar y diagnostico de maquinas rotatorias es muy comun emplear metodos de calculo sencillos para el analisis de esfuerzos mecanicos y termicos, analisis dinamico y termodinamico y de flujo de fluidos en estas maquinas (Gutierrez et al., 1989). El analisis con estos metodos proporcionan los resultados necesarios para la etapa del proyecto inicial de la maquina. Posteriormente, para refinar el diseno de algunos componentes de la maquina, se aplican las herramientas mas complejas. En el informe de Gutierrez et al., (1989) se desarrollan 34 programas para el diseno preliminar y diagnostico de equipo rotatorio; en este articulo, se presenta uno de ellos, en el que se emplea un metodo para el analisis de esfuerzos mecanicos y termicos en discos de espesor constante o variable que se encuentran comunmente en turbomaquinas y en equipos rotatorios.

  5. Computer program for the calculation of stresses in rotary equipment discs; Programas de computo para el calculo de esfuerzos en discos de equipo rotatorio

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez Delgado, Wilson; Kubiak, Janusz; Serrano Romero, Luis Enrique [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1991-12-31

    In the preliminary design and diagnosis of rotary machines is very common to utilize simple calculation methods for the mechanical and thermal stresses, dynamic and thermodynamic analysis and flow of fluids in this machines (Gutierrez et al., 1989). The analysis with these methods provides the necessary results for the project initial stage of the machine. Later on, more complex tools are employed to refine the design of some machine components. In the Gutierrez report et al., (1989) 34 programs were developed for the preliminary design and diagnosis of rotating equipment; in this article, one of them is presented in which a method for the analysis of mechanical and thermal stresses is applied in discs of uniform or variable thickness that are normally found in turbomachines and rotary equipment. [Espanol] En el diseno preliminar y diagnostico de maquinas rotatorias es muy comun emplear metodos de calculo sencillos para el analisis de esfuerzos mecanicos y termicos, analisis dinamico y termodinamico y de flujo de fluidos en estas maquinas (Gutierrez et al., 1989). El analisis con estos metodos proporcionan los resultados necesarios para la etapa del proyecto inicial de la maquina. Posteriormente, para refinar el diseno de algunos componentes de la maquina, se aplican las herramientas mas complejas. En el informe de Gutierrez et al., (1989) se desarrollan 34 programas para el diseno preliminar y diagnostico de equipo rotatorio; en este articulo, se presenta uno de ellos, en el que se emplea un metodo para el analisis de esfuerzos mecanicos y termicos en discos de espesor constante o variable que se encuentran comunmente en turbomaquinas y en equipos rotatorios.

  6. Thermodynamic study of residual heat from a high temperature nuclear reactor to analyze its viability in cogeneration processes; Estudio termodinamico del calor residual de un reactor nuclear de alta temperatura para analizar su viabilidad en procesos de cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Santillan R, A.; Valle H, J.; Escalante, J. A., E-mail: santillanaura@gmail.com [Universidad Politecnica Metropolitana de Hidalgo, Boulevard acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2015-09-15

    In this paper the thermodynamic study of a nuclear power plant of high temperature at gas turbine (GTHTR300) is presented for estimating the exploitable waste heat in a process of desalination of seawater. One of the most studied and viable sustainable energy for the production of electricity, without the emission of greenhouse gases, is the nuclear energy. The fourth generation nuclear power plants have greater advantages than those currently installed plants; these advantages have to do with security, increased efficiencies and feasibility to be coupled to electrical cogeneration processes. In this paper the thermodynamic study of a nuclear power plant type GTHTR300 is realized, which is selected by greater efficiencies and have optimal conditions for use in electrical cogeneration processes due to high operating temperatures, which are between 700 and 950 degrees Celsius. The aim of the study is to determine the heat losses and the work done at each stage of the system, determining where they are the greatest losses and analyzing in that processes can be taken advantage. Based on the study was appointed that most of the energy losses are in form of heat in the coolers and usually this is emitted into the atmosphere without being used. From the results a process of desalination of seawater as electrical cogeneration process is proposed. This paper contains a brief description of the operation of the nuclear power plant, focusing on operation conditions and thermodynamic characteristics for the implementation of electrical cogeneration process, a thermodynamic analysis based on mass and energy balance was developed. The results allow quantifying the losses of thermal energy and determining the optimal section for coupling of the reactor with the desalination process, seeking to have a great overall efficiency. (Author)

  7. Helium desorption in EFDA iron materials for use in nuclear fusion reactors; Desorcion de helio en materiales de fierro EFDA para su aplicacion en los reactores de fusion nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Salazar R, A. R.; Pinedo V, J. L. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sanchez, F. J.; Ibarra, A.; Vila, R., E-mail: arsr2707@hotmail.com [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Av. Complutense No. 40, 28040 Madrid (Spain)

    2015-09-15

    In this paper the implantation with monoenergetic ions (He{sup +}) was realized with an energy of 5 KeV in iron samples (99.9999 %) EFDA (European Fusion Development Agreement) using a collimated beam, after this a Thermal Desorption Spectrometry of Helium (THeDS) was made using a leak meter that detects amounts of helium of up to 10{sup -}- {sup 12} mbar l/s. Doses with which the implantation was carried out were 2 x 10{sup 15} He{sup +} /cm{sup 2}, 1 x 10{sup 16} He{sup +} /cm{sup 2}, 2 x 10{sup 16} He{sup +} /cm{sup 2}, 1 x 10{sup 17} He{sup +} /cm{sup 2} during times of 90 s, 450 s, 900 s and 4500 s, respectively. Also, using the SRIM program was calculated the depth at which the helium ions penetrate the sample of pure ion, finding that the maximum distance is 0.025μm in the sample. For this study, 11 samples of Fe EFDA were prepared to find defects that are caused after implantation of helium in order to provide valuable information to the manufacture of materials for future fusion reactors. However understand the effects of helium in the micro structural evolution and mechanical properties of structural materials are some of the most difficult questions to answer in materials research for nuclear fusion. When analyzing the spectra of THeDS was found that five different groups of desorption peaks existed, which are attributed to defects of He caused in the material, these defects are He{sub n} V (2≤n≤6), He{sub n} V{sub m}, He V for the groups I, II and IV respectively. These results are due to the comparison of the peaks presented in the desorption spectrum of He, with those of other authors who have made theoretical calculations. Is important to note that the thermal desorption spectrum of helium was different depending on the dose with which the implantation of He{sup +} was performed. (Author)

  8. Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2007-07-01

    Presently work the {beta} and {gamma} radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of {beta} and {gamma} radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the {gamma} radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U. [IPEN, CNEN-SP, P.O. Box 11049, CEP 05422-970, Sao Paulo (Brazil); Riella, H. G. [Universidade Federal de Santa Catarina, Departamento de Engenharia Quimica, Campus Universitario, Florianopolis, CEP 88040-900 (Brazil)]. e-mail: grsantos@ipen.br

    2008-07-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% {sup 2}35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

  14. 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

  15. 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

  16. Partida de um reator anaeróbio horizontal para tratamento de efluentes do processamento dos frutos do cafeeiro Start-up of an anaerobic horizontal-flow reactor for treating wastewater from a coffee fruits processing

    Directory of Open Access Journals (Sweden)

    Alisson C. Borges

    2009-01-01

    Full Text Available O presente estudo teve o objetivo de avaliar a partida e a adaptação de um reator anaeróbio horizontal de leito fixo (RAHLF no tratamento de águas residuárias do processamento primário dos frutos do cafeeiro (ARC. O reator foi construído com tubos de PVC de 0,2 m de diâmetro e 3,2 m de comprimento. O sistema foi preenchido com cubos de espuma de poliuretano para imobilização de biomassa ativa. O reator apresentou volume total de 0,1 m³ e volume útil equivalente a 0,04 m³. Em média, houve remoção de 49% da matéria orgânica, com o reator trabalhando sob carga orgânica volumétrica média de 2,66 kg m-3 d-1, medida como DQO. A suplementação de alcalinidade, somada à inoculação prévia de biomassa, proporcionou partida estável do RAHLF, confirmada pelo consumo de ácidos voláteis e adaptação da microbiota ao resíduo. O sistema apresentou resistência às variações de vazão e de carga orgânica observadas, e os teores de fenol e potássio monitorados não causaram inibição da atividade biológica no RAHLF. O maior controle sobre as variações de carga é fator importante na continuidade dos estudos.This study aimed to evaluate the start-up and the adaptation of an anaerobic horizontal-flow immobilized biomass (HAIB reactor in order to treat wastewater from a primary processing of coffee fruits. The reactor was built with PVC tubes of 0.2 m in diameter and 3.2 m in length. The system was filled with cubes of polyurethane foam for immobilization of active biomass. The reactor presented a total capacity of 0.1 m³ and reaction volume equal to 0.04 m³. 49% of organic matter. Removal efficiency was observed, with medium organic volumetric loads equal to 2.66 kg m-3 d-1 (as chemical oxygen demand. The supplementary addition of alkalinity and the previous biomass inoculation provided a stable start-up of the reactor, as confirmed by the reduction of volatile acids and an adaptation of the present microbiology community

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  18. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  19. 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

  20. 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

  1. 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.)

  2. 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.)

  3. 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)

  4. 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

  5. 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

  6. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor; Analisis para el acoplamiento del codigo NESTLE para la cinetica tridimensional del nucleo al codigo avanzado de sistemas termo-hidraulicos, RELAP5/SCDAPSIM y su aplicacion al reactor de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Salazar C, J H; Nunez C, A [CNSNS, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D.F. (Mexico); Chavez M, C [UNAM, Facultad de Ingenieria, DEPFI Campus Morelos (Mexico)

    2004-07-01

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  7. 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

  8. Statistical analysis in the design of nuclear fuel cells and training of a neural network to predict safety parameters for reactors BWR; Analisis estadistico en el diseno de celdas de combustible nuclear y entrenamiento de una red neuronal para predecir parametros de seguridad para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jauregui Ch, V.

    2013-07-01

    In this work the obtained results for a statistical analysis are shown, with the purpose of studying the performance of the fuel lattice, taking into account the frequency of the pins that were used. For this objective, different statistical distributions were used; one approximately to normal, another type X{sup 2} but in an inverse form and a random distribution. Also, the prediction of some parameters of the nuclear reactor in a fuel reload was made through a neuronal network, which was trained. The statistical analysis was made using the parameters of the fuel lattice, which was generated through three heuristic techniques: Ant Colony Optimization System, Neuronal Networks and a hybrid among Scatter Search and Path Re linking. The behavior of the local power peak factor was revised in the fuel lattice with the use of different frequencies of enrichment uranium pines, using the three techniques mentioned before, in the same way the infinite multiplication factor of neutrons was analyzed (k..), to determine within what range this factor in the reactor is. Taking into account all the information, which was obtained through the statistical analysis, a neuronal network was trained; that will help to predict the behavior of some parameters of the nuclear reactor, considering a fixed fuel reload with their respective control rods pattern. In the same way, the quality of the training was evaluated using different fuel lattices. The neuronal network learned to predict the next parameters: Shutdown Margin (SDM), the pin burn peaks for two different fuel batches, Thermal Limits and the Effective Neutron Multiplication Factor (k{sup eff}). The results show that the fuel lattices in which the frequency, which the inverted form of the X{sup 2} distribution, was used revealed the best values of local power peak factor. Additionally it is shown that the performance of a fuel lattice could be enhanced controlling the frequency of the uranium enrichment rods and the variety of

  9. A microwave technique for electrical conductivity measurements in semiconductors; Un procedimiento en frecuencia de microondas para la medicion de la conductividad electrica en semiconductores

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez, A; Zehe, A [Benemerita Universidad Autonoma de Puebla, Puebla (Mexico); Starostenko, O [Universidad de las Americas, Puebla, (Mexico); Perez, L [Benemerita Universidad Autonoma de Puebla, Puebla (Mexico)

    2003-01-15

    In the present paper a theoretical approach is given together with the description of the experimental technique for electrical conductivity measurements of semiconductors in frequency range of 10 GHz, corresponding to a wavelength of {lambda}= 3 cm. It is shown, that the sample conductivity, measured without the need of electrical contacts, Is practically identical with results obtained by direct current methods. The potential use of the method for dielectric measurements in samples of general shape is included in the discussion. Given that the local electrical field is known only in bodies of ellipsoidal shape, one has to apply approximation methods for sample shapes of practical relevance (cylinders, cubes, disks). Finally, the measuring range and corresponding errors are explained on the base of measurements realized with silicon samples. [Spanish] En el presente trabajo se desarrolla la teoria junto con la descripcion del arreglo experimental para la medicion de la conductividad electrica en muestras semiconductoras, utilizando un campo de microondas de {lambda}= 3 cm, correspondiente a una frecuencia n de 10 GHz. Se demuestra que la conductividad electrica medida sin la necesidad de contactos electricos es practicamente identica con resultados obtenidos por el metodo comun utilizando corriente directa (d.c.). Se incorpora en la discusion ademas el potencial del metodo para el estudio de propiedades dielectricas en muestras de geometria general. Dado que solamente en cuerpos con geometria elipsoidal se conoce bien el campo electrico local, y con esto el momento bipolar inducido, para muestras de geometrias mas practicas (cilindros, cubos, discos) se acude a metodos de aproximacion. Finalmente, se discute el rango de mediciones y errores, y se presentan mediciones concretas, utilizando muestras de silicio.

  10. Métodos cromatográficos para determinar aminas biogênicas em alimentos de origem animal

    Directory of Open Access Journals (Sweden)

    César Aquiles Lázaro de la Torre

    2013-12-01

    Full Text Available Aminas biogenicas sao formadas como resultado da descarboxilacao de aminoacidos livres especificos. A analise desses metabolitos e de grande importancia na determinacao da qualidade e monitoramento de biogenicas como histamina e tiramina relacionadas com episodios de intoxicacao em humanos. A cromatografia e uma tecnica de separacao química usada para caracterizar aminas biogenicas. Variacoes da tecnica (cromatografia liquida, em camada delgada e gasosa tem sido amplamente usadas, porem a complexidade da matriz alimentar faz com que sejam realizadas mudancas nos processos de extracao, derivatizacao e deteccao em concordancia com cada grupo de alimento. A cromatografia liquida de alta eficiencia (CLAE e o metodo mais utilizado na determinacao de aminas biogenicas em alimentos. Contudo, devido a importancia das aminas biogenicas no controle da qualidade e a seguranca do consumidor, os pesquisadores tentam desenvolver novos metodos com o intuito de uma analise mais rapida e precisa para o controle de alimentos no mercado. O objetivo da revisao e apresentar algumas tecnicas cromatograficas aplicadas no monitoramento de aminas biogenicas em produtos de origem animal.

  11. 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.)

  12. 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

  13. 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

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  15. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  16. COMPARACIÓN DEL ESTADIO FETAL OBTENIDO POSTMORTEM MEDIANTE DOS METODOS ANTROPOMETRICOS. Comparación del estadio fetal obtenido postmortem mediante dos metodos antropometricos

    Directory of Open Access Journals (Sweden)

    Mirta M Aliendo

    2016-03-01

    Full Text Available En la etapa fetal se observa un rápido incremento de la masa corporal y de todas las dimensiones. La literatura evidencia discrepancias sobre los criterios para determinar el estadio fetal post-mortem, en relación a los parámetros morfométricos utilizados, por lo que nuestro objetivo fue comparar la medición morfológica directa (vertex-coxis, tabla de Hansmann con la ultrasonografía (medición del fémur, para establecer el grado de confiabilidad en la determinación post-mortem del estadío fetal. Se utilizaron 120 fetos: 1 grupo A (60 fetos estadificado ecográficamente y 2 grupo B (60 fetos estadificado por tabla de Hansmann. A ambos grupos se le realizaron múltiples mediciones siguiendo parámetros probados según la literatura internacional. Se utilizó calibre de precisión. Parámetros evaluados: longitud vertex-coxis, circunferencia cefálica, diámetro cefálico occipito-frontal, biparietal, longitud mentón-vertex, perímetro toráxico-transverso, circunferencia abdomi-nal y longitudes de brazo, antebrazo, mano, muslo, pierna y pie. Estos valores fueron agrupados por semanas, obteniéndose la media y aplicándose la prueba t de Student. Los resultados demostraron que la diferencia entre los parámetros medidos en el grupo A y en el grupo B eran significativas en todas las semanas, por lo que se observa disparidad en la determinación del estadio fetal por ecografía y los registros correspondientes a la medición vertex-coxis (tabla de Hansmann postmortem. Concluímos que los resultados obtenidos por ambas modalidades de medición son diferentes para una misma edad gestacional y, por ende, resultaría más apropiado referirse a fetos con ciertas dimensiones según alguno de estos parámetros que a “edad gestacional”.In fetal stage, body mass and measurements quickly increase. Scientific literature shows differences on the criteria to determine the post-mortem fetal stage, depending on morphometric parameters. Our objective

  17. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors; Desarrollo de un program de computo de calculo rapido para el prediseno de celdas de combustible nuclear avanzado 10 x 10 para reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia, R.; Montes, J.L.; Ortiz, J.J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2005-07-01

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  18. Comparative analysis to determine asphalt density and content, using nuclear and traditional methods; Analisis comparativo para la determinacion de densidad y contenido de asfalto, mediante metodo nuclear y metodo tradicional

    Energy Technology Data Exchange (ETDEWEB)

    Margffoy S, F R; Robayo S, A M

    1989-07-01

    Quality control for flex pavement construction in Colombia for asphaltic mix, as well as for granular and granular sub base layers is made by means of methods that does not guarantee the quality of the job, due to the difficult execution of tests, which impede more to be done or due to inherent problems of the test process. Thanks to the inherent characteristics and advantages of nuclear techniques, those become the optimal alternative to this quality control job. The present research project has been developed with the objective of justifying the use of new technologies applied to road construction; making a comparative analysis between traditional methods used in our country and nuclear techniques that have been using in United States with great success in quality control in road construction.

  19. The disc method. A new method for selecting facilitations in flocculating sludge to be dewatered in centrifuges; Metodo de disco. Un nuevo metodo para la seleccion de floculantes en la floculacion de lodos a deshidratar en centrifugas

    Energy Technology Data Exchange (ETDEWEB)

    Canga Rodriguez, J.; Gutierrez Lavin, A.

    2002-07-01

    An experimental protocol was designed at a laboratory scale, in view of achieving the selection with different poly electrolytes related to the chemical conditioning (flocculation) of sewage sludge before dewatering it in a drying centrifuge. The method is based on a new parameter of quality of the formed floc, which measures its compaction when is submitted to a fix external strength. Some experimental tests have been introduced, whose results are numbers, avoiding all subjective aspects related to direct observation of flocs. (Author) 8 refs.

  20. Analysis of the indices of thermal comfort for the conditions of the Mexican Republic; Analisis de los indices de confort termico para las condiciones de la republica mexicana

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes Freixanet, Victor; Rodriguez Viqueira, Manuel [Universidad Autonoma Metropolitana - Unidad Azcapotzalco (Mexico)

    2009-07-15

    The objective of this article is to analyze different indices of thermal comfort for the Mexican Republic. Among them the Fanger (PMV and PPD) physiological methods of comfort and the new effective temperature index are included. The standard effective temperature (SET), as well as the adaptive methods of Humphreys and Nicol, Auliciems, De Dear and Brager. A comparative analysis is done of the different indices through thematic maps determined by interpolation, using a climatic data base of 700 cities obtained from the observatories and stations of the National Meteorological Service. This article pretends to establish general criteria of the thermal comfort to later define design strategies for each one of the climatic regions of the Mexican Republic. [Spanish] El objetivo de este articulo es analizar distintos indices de confort termico para la Republica Mexicana. Entre ellos se incluyen los metodos fisiologicos de confort de Fanger (PMV y PPD), el indice de nueva temperatura efectiva. La temperatura efectiva estandar (SET), asi como los metodos adaptativos de Humphreys y Nicol, Auliciems, De Dear y Brager. Se hace un analisis comparativo de los distintos indices a traves de mapas tematicos determinados por interpolacion, usando una base de datos climaticos de 700 ciudades obtenidos de los observatorios y estaciones del Servicio Meteorologico Nacional. Este articulo presenta establecer criterios generales del confort termico para posteriormente definir estrategias de diseno para cada una de las regiones climaticas de la Republica Mexicana.

  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. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  4. 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.)

  5. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  6. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  7. SOLUCIÓN ANALÍTICA PARA OBTENER EL VOLUMEN ÓPTIMO DE UNA SERIE DE REACTORES DE AGITACIÓN CONTINUA DONDE SE EFECTÚA UNA REACCIÓN DE PRIMER ORDEN

    Directory of Open Access Journals (Sweden)

    Ignacio Elizalde

    2013-01-01

    Full Text Available An analytical procedure for determining the optimum size of CSTR in series operating under isothermal and isobaric conditions sustaining first order reaction at constant density has been developed. The procedure requires the concentration of reactant at the entrance of the first reactor and at the outlet of the last reactor; it is also required the continuity of reaction rate as function of conversion, due to the later changes from one reactor to another. The optimization method involves the calculation of intermediate concentrations instead of their estimation, as it is done by graphical solution reported previously. Also, the procedure reported in this contribution is valid for any reactor number. Under these circumstances the method predicts that all reactors must have the same size in order to minimize the total volume of the system.

  8. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  9. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  10. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  12. Uses and updating of the Benders method in the integer-mixed programming in the planning of the electric power systems expansion; Usos y actualizacion del metodo de Benders en la programacion entera-mixta y en la planeacion de la expansion de los sistemas electricos de potencia

    Energy Technology Data Exchange (ETDEWEB)

    De la Torre Vega, Eli

    1997-04-01

    of some potential improvements that the proposed method has. [Espanol] En el primer capitulo se presenta la deduccion de los cortes de Benders partiendo de las propiedades de la dualidad. Tambien se presentan las propiedades de los cortes de Benders asi como el algoritmo inicial de Benders para resolver cualquier problema de programacion lineal entera-mixta. En el segundo capitulo, se presenta el problema de la planeacion de la expansion de los medios de generacion y transmision en un sistema electrico de potencia, las distintas estructuras de la programacion matematica a que da lugar y como se puede adaptar el metodo de Benders a estas. En el tercer capitulo se presentan las aportaciones teoricas de este trabajo: a) Como inicializar el problema maestro para aprovechar la experiencia adquirida despues de haber resuelto un problema similar, de modo que se pueda resolver mas eficientemente, la sucesion de problemas de programacion lineal entero-mixtos que surgen al resolver el problema de la planeacion de la expansion de los medios de generacion y transmision en un sistema electrico de potencia. b) Como generar un problema maestro cuya solucion optima continua corresponda al optimo continuo del problema entero-mixto, de modo que la busqueda de soluciones enteras se realice en la vecindad del optimo continuo. c) Como generar una solucion entera, cercana al optimo continuo del problema entero-mixto, que tenga mucha probabilidad de ser factible, y que quizas sea la solucion optima entera, en un tiempo menor al que se requiere para resolverlo en forma exacta. Ademas, se presentan otras ideas que se le pueden incorporar al metodo de Benders. A fin de mostrar la efectividad de las ideas propuestas, en el capitulo 4 se presentan los resultados obtenidos al resolver varios problemas usando: 1. El metodo de Benders in actualizacion, 2. El metodo de Ramificacion y Acotamiento, 3. La actualizacion de Benders al agregar restricciones y 4. La actualizacion de Benders al considerar

  13. 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.)

  14. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  15. Gas-flow detector for uranium contamination on finned-can surface of a reactor fuel; Detecteur a courant gazeux pour deceler la contamination en uranium des nervures des gaines de combustible nucleaire; Gazopotochnyj detektor zagryazneniya uranom rebristoj poverkhnosti obolochki reaktornykh teplovydelyayushchikh ehlementov; Detector de flujo gaseoso para medir la contaminacion de uranio en la superficie de la vaina de aletas de los elementos combustibles para reactores

    Energy Technology Data Exchange (ETDEWEB)

    Miwa, H; Shiojiri, T; Maeda, Y [Kobe Kogyo Corporation, Okubo, Akashi, Hyogo (Japan)

    1962-04-15

    alpha. Avant chaque mesure, le compteur est vide au moyen d'une pompe rotative; on y admet ensuite du gaz PR (melange de 90% d'argon et de 10% de methane). En utilisant ce nouvel appareil, les auteurs ont reussi a deceler les particules alpha emises par 1 x 10{sup -5} gramme d'uranium naturel contaminant les nervures des gaines de combustible d'unreacteur du type Calder Hall; le combustible du reacteur de recherche japonais JRR-3 sera inspecte a l'aide de ce compteur. (author) [Spanish] El detector de corriente gaseosa descrito es un contador proporcional de rejilla especialmente destinado a determinar la contaminacion del uranio en la superficie de la vaina de aletas de los elementos combustibles para reactores. Con el tipo comun de contador proporcional, compuesto solamente de catodo y colector, apenas es posible descubrir las particulas a emitidas por el uranio contaminado en superficies irregulares tales como las aletas de la envoltura de un elemento combustible, debido a la falta de uniformidad del campo electrico en 'as cercanias de la superficie. Este es el motivo que indujo a los autores a construir el contador proporcional de rejilla. Este contador consta del elemento combustible, una rejilla, unos colectores y un catodo de forma cilindrica dispuestos coaxialmente. El combustible va colocado en el centro de la rejilla y se aplica una tension negativa. El espacio entre el combustible y la rejilla actua como camara colectora de iones. La rejilla esta formada por delgados alambres de tungsteno paralelos dispuestos cilindricamente en torno del elemento combustible y conectados a masa. Los colectores son 16 alambres finos de tungsteno de construccion semejante a la de la rejilla, pero cada uno de los alambres esta electricamente aislado de los restantes. Todos los colectores estan conectados entre si a traves de resistencias de 50 k{Omega} y conectados tambien a una fuente positiva de alta tension a traves de una resistencia. El espacio entre la rejilla, los

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  19. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  20. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  1. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  2. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor; Modelo simplificado para simulacao do comportamento termohidraulico do canal quente de reator nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Belem, J A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs.

  3. Neutron Investigation of Magnon Spectrum in Haematite; Etude du Spectre de Magnons dans l'Hematite, au Moyen des Neutrons; Nejtronnoe issledovanie spektra magnona v gematite; Estudio, por Metodos Neutronicos, del Espectro de Magnones en la Hematita

    Energy Technology Data Exchange (ETDEWEB)

    Dimitrijevic, Z.; Rzany, H.; Todorovic, J.; Wanic, A. [Institute for Nuclear Physics Cracow (Poland)

    1965-04-15

    'existence d'une branche optique qui, toutefois, ne se prete pas aux recherches par la methode de la diffusion des neutrons. (author) [Spanish] Los autoresrealizaron sus mediciones en el reactor RA de Vinca con ayuda de un espectrometro neutronicode cristal. Los neutrones monocromaticos ({lambda} = 1.314A) se dispersaron en un monocristal de hematita (Fe{sub 2}O{sub 3}{alpha}) de gran tamano. Se estudiaron las distribuciones angulares de los neutrones inelast Lea mente dispersos, midiendo para un cierto numero de angulos distintos de desplazamiento del cristal, {Delta}{theta}, la anchura {Gamma} del haz de neutrones dispersos (el llamado cono de dispersion), que fue adscrito a la superficie de dispersion magnOnica que envolvia al punto(l,l,l)delared reciproca. Se calcularon y compararon con los puntos obtenidos experimentalmente los valores de {Gamma} en funcion de la variacion del angulo de desplazamiento para diversas velocidades de los magnones. Se comprobo que el valor de la velocidad en la direccion [111] era igual a 25,5 {+-} 1,0 km/s, descubriendose asi la anisotropia estructural de la relacion de dispersion magnonica. Se encontro que la velocidad era superior para las direcciones de propagacion paralelas al eje [111], lo que esta cualitativamente de acuerdo con anteriores mediciones (Riste y otros) que daban una velocidad v = 38 km/s. No fue posible poner de manifiesto la existencia de una laguna energetica Eg en la banda de las energias magnonicas acusticas. Se calculo que el valor de Eg tiene que ser inferior a 1 meV. Utilizando las formulas propuestas por Wallace se calcularon las relaciones de dispersion de los magnones en la hematita. Este calculo se efectuo en el supuesto de que existan dos integrales de intercambio que no desaparecen: J{sub 1} y J{sub 2}, que indican el acoplamiento energetico de superintercambio entre los spins de los iones hierro proximos, enlazados por un ion oxigeno, los angulos de los enlaces Fe-O-Fe tenidos en cuenta son de 132 Degree

  4. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  6. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  7. Development of a protection system for research reactor based in Field Programmable Gate Array - FPGA; Desenvolvimento de sistema de protecao para reator nuclear de pesquisa baseado em Field Programmable Gate Array - FPGA

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Roque Hudson da Silva

    2016-07-01

    This study presents a implementation purpose of a protection system for research nuclear reactors by using a programed device FPGA (Field Programmable Gate Array). As well as logic protection method involved on an automatic shutdown (TRIP) of a reactor, that ensure the security on such systems. These new control and operation mechanics are developed to guarantee that the security limits of a power plant are not exceeded, these mechanics can work isolated or in groups to safe guard the security levels. For this implementation to be completed, there will be presented the main aspects and concepts referred to protection systems, mostly about research nuclear reactors, with some applications terms exposed. The system proposed at this paper was developed following the VHDL (Very High Speed Integrated Circuits) hardware describing language, and the Modelsim software from Altera Software to program the automatic turning off routines, and hypothetical simulations for such. The results show that for every software application for supporting nuclear reactors, like security devices, they have to meet the IEC 60880 criteria. This paper have great importance, seeing that nuclear reactor security systems, are a basic element for ensure the reactor security. (author)

  8. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  9. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  10. A pulsed fast reactor; Un reacteur pulse a neutrons rapides; Impul'snyj reaktor na bystrykh nejtronakh; Reactor rapido pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, G. E.; Blokhintsev, D. I.; Blyumkina, Yu. A.; Bondarenko, I. I.; Deryagin, B. N.; Zajmovskij, A. S.; Zinov' ev, V. P.; Kazachkovskij, O. D.; Krasnoyarov, N. V.; Lejpunskij, A. I.; Malykh, V. A.; Nazarov, P. M.; Nikolaev, S. K.; Stavisskij, Yu. Ya.; Ukraintsev, F. I.; Frank, I. M.; Shapiro, F. Ji.; Yazvitskij, Yu. S. [Akademiya Nauk, Moscow, SSSR (Russian Federation)

    1962-03-15

    temperature de la reactivite. (author) [Spanish] Desde diciembre de 1960, el reactor de impulsos de neutrones rapidos IBR viene funcionando a su potencia nominal en el Instituto Central de Investigaciones Nucleares. Dicho reactor se utiliza como fuente pulsante de neutrones para realizar experimentos de fisica por el metodo del tiempo de vuelo. Se llevan a cabo determinacione s de las secciones eficaces totales, de las secciones eficaces de captura de neutrones intermedios, estudios de las interacciones de los neutrones lentos con los solidos y los liquidos y mediciones de los espectros neutronicos en distintos medios. Los autores describen las principales caracteristicas constructivas del reactor y los resultados de los estudios realizados mediante el mismo. Este reactor trabaja con arreglo a un regimen de impulsos periodicos. Los impulsos de potencia se originan cuando la parte movil del cuerpo, fijada a un disco giratorio, atraviesa la parte estacionaria con una velocidad del orden de los 230 m/s. Gracias a la presencia de una zona movible auxiliar, es posible variar la frecuencia de los impulsos de potencia entre 2,3 y 88 impulsos por segundo. La potencia media del reactor es de 1 kW y la duracion media de los impulsos, de 36 {mu}s. El reactor esta provisto de un sistema de mando y de seguridad que vela por el mantenimiento automatico de la potencia del reactor en su valor medio, asi como por su rapida detencion en caso de perturbacion del funcionamiento. Tambien posee el reactor conductores neutronicos de vacio, que se utilizan en los experimentos de tiempo de vuelo. El conducto principal tiene 1000 m de longitud. En el proceso de puesta en marcha y durante las investigaciones fisicas realizadas con el reactor, se estudio el efecto del desplazamiento de los organos de regulacion y de las partes moviles del cuerpo sobre la reactividad, se determino la duracion de los impulsos a distintos regimenes de trabajo del reactor y se estudiaron las fluctuaciones de la amplitud de

  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. 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

  13. Application of autoregressive methods and Lyapunov coefficients for instability studies of nuclear reactors; Aplicação de métodos autorregressivos e coeficientes de Lyapunov para estudos de instabilidades em reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Aruquipa Coloma, Wilmer

    2017-07-01

    Nuclear reactors are susceptible to instability, causing oscillations in reactor power in specific working regions characterized by determined values of power and coolant mass flow. During reactor startup, there is a greater probability that these regions of instability will be present; another reason may be due to transient processes in some reactor parameters. The analysis of the temporal evolution of the power reveals a stable or unstable process after the disturbance in a light water reactor of type BWR (Boiling Water Reactor). In this work, the instability problem was approached in two ways. The first form is based on the ARMA (Autoregressive Moving Average models) model. This model was used to calculate the Decay Ratio (DR) and natural frequency (NF) of the oscillations, parameters that indicate if the one power signal is stable or not. In this sense, the DRARMA code was developed. In the second form, the problems of instability were analyzed using the classical concepts of non-linear systems, such as Lyapunov exponents, phase space and attractors. The Lyapunov exponents quantify the exponential divergence of the trajectories initially close to the phase space and estimate the amount of chaos in a system; the phase space and the attractors describe the dynamic behavior of the system. The main aim of the instability phenomena studies in nuclear reactors is to try to identify points or regions of operation that can lead to power oscillations conditions. The two approaches were applied to two sets of signals. The first set comes from signals of instability events of the commercial Forsmark reactors 1 and 2 and were used to validate the DRARMA code. The second set was obtained from the simulation of transient events of the Peach Bottom reactor; for the simulation, the PARCS and RELAP5 codes were used for the neutronic/thermal hydraulic coupling calculation. For all analyzes made in this work, the Matlab software was used due to its ease of programming and

  14. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  15. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  16. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  18. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  19. Measurements with a Pulsed and Modulated Source in a Reactor; Mesures au Moyen d'une Source Pulsee et Modulee dans un Reacteur; Izmereniya v reaktore s pomoshch'yu impul'snogo i moduliruemogo is tochnika; Mediciones Efectuadas en Reactor con una Fuente Pulsada y Modulada

    Energy Technology Data Exchange (ETDEWEB)

    Rotter, W. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

    1965-10-15

    generateur: le temps de mesure est donc minimum. Les observations enregistrees sur bande perforee sont depouillees par une calculatrice numerique. (author) [Spanish] Los laboratorios de investigacion Philips han construido un generador neutronico de flujo variable en funcion del tiempo. Con una serie de mediciones efectuadas en el reactor BRO2 en estado subcrftico, se ha demostrado su utililidad practica en la esfera de la fisica de los reactores. El funcionamiento del generador es muy flexible debido a su alta estabilidad, a la posibilidad de variar bruscamente la intensidad neutronica, y de pulsar el flujo o modularlo de manera sinusoidal. El generador permite determinar la reactividad ({rho} = {Delta}k/{beta}) y la vida media de los neutrones ( Script-Small-L /{beta}) segun varios metodos independientes. Es posible proceder a una comparacion exacta de esos metodos, dado que pueden aplicarse sin modificar las condiciones de medicion. El autor ha calculado los siguientes valores: a) p, sobre la base de los neutrones retardados, por reduccion instantanea del flujo neutronico; b) p, sobre la base de los neutrones inmediatos, por impulsos neutronicos; c) Script-Small-L /{beta}, combinando 1) y 2), cuando 0, 5 dolares < {rho} < 2 dolares, y d) Script-Small-L /{beta}, sobre la base de la funcion de transferencia del reactor para una fuente modulada. En la memoria se examinan las funciones de transferencia correspondientes a un oscilador de reactividad y a una fuente de modulacion sinusoidal. Se demuestra que es posible medir Script-Small-L /{beta}, cuando 0,1 dolar < {rho} < 10 dolares utilizando una fuente modulada. Por el mismo metodo se obtiene tambien la reactividad partiendo de la razon neutrones inmediatos/neutrones retardados para una frecuencia optima que es practicamente independiente de los datos relativos a los neutrones retardados y del cociente Script-Small-L /{beta}. La precision estadistica de cada metodo puede aumentarse acumulando un gran numero de ciclos en el

  20. 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

  1. Evaluación de una prueba rápida para el diagnostico de Malaria en áreas endémicas del Perú

    Directory of Open Access Journals (Sweden)

    Nancy Arrospide V

    2006-04-01

    Full Text Available Objetivos: Evaluar la sensibilidad, especificidad y valores predictivos de la prueba rapida basada en la deteccion de la pLDH (OptiMALR kit individual para el diagnostico de malaria en areas endemicas del Peru. Materiales y metodos: Estudio transversal realizado con pacientes febriles atendidos en centros de salud de la selva norte del Peru (San Martin y Loreto, de abril a diciembre de 2001. A cada paciente se le realizo la gota gruesa, la prueba OptiMALR y densidad parasitaria en forma ciega, por personal local capacitado y luego en el Laboratorio Nacional de Referencia de Malaria. Se calculo la sensibilidad (S, especificidad (E, valor predictivo positivo (VPP y valor predictivo negativo(VPN de la prueba OptiMALR en relacion a la gota gruesa para el diagnostico de malaria global y segun especie (P.falciparum y P.vivax. Resultados: Se incluyeron 346 muestras, 170 positivas. La prueba OptiMALR tuvo niveles de S=95,7%, E=97,1%, VPP=97,7%, VPN=95,3% independientemente de la especie. Para P.falciparum tuvo S=90,5%, E=97,3%, VPP=67,9 y VPN=99.4%; en tanto que para P.vivax S=92,0%, E=99,0%, VPP=98,7% y VPN=93,5%. Las sensibilidades estratificadas por parasitemia fueron 97,0% (5000 parasitos/¥ìL, 99% (100-5000 p/¥ìL y 50% (<100p/¥ìL. Conclusiones: La prueba rapida OptiMALR es un metodo con buena sensibilidad y especificidad para el diagnostico de malaria y puede ser usado en lugares donde no se dispone de laboratorios o microscopistas.

  2. Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I; Comparacion entre un modelo de diferencias finitas (PUMA) y uno de elementos finitos (DELFIN) para la simulacion del reactor de la CNA-I (central nuclear Atucha-I)

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C R [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    The reactor code PUMA, developed in CNEA, simulates nuclear reactors discretizing space in finite difference elements. Core representation is performed by means a cylindrical mesh, but the reactor channels are arranged in an hexagonal lattice. That is why a mapping using volume intersections must be used. This spatial treatment is the reason of an overestimation of the control rod reactivity values, which must be adjusted modifying the incremental cross sections. Also, a not very good treatment of the continuity conditions between core and reflector leads to an overestimation of channel power of the peripherical fuel elements between 5 to 8 per cent. Another code, DELFIN, developed also in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and current among elements and a more realistic representation of the hexagonal lattice of the reactor. A comparison between results obtained using both methods in done in this paper. (author). 4 refs., 3 figs.

  3. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  4. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  5. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  7. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  9. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  10. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  11. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  12. Produção de biogás no tratamento dos efluentes líquidos do processamento de Coffea arabica L. em reator anaeróbico UASB para o potencial aproveitamento na secagem do café Biogas production in the treatment of Coffea arabica L. processing wastewaters in UASB anaerobic reactor for the potential use in the coffee drying

    Directory of Open Access Journals (Sweden)

    Marco Antônio Calil Prado

    2008-06-01

    Full Text Available Estudou-se a produção de biogás proveniente do tratamento das águas residuárias do processamento por via úmida do café (ARC coco em sistema de tratamento anaeróbio em escala laboratorial. O sistema foi composto de um tanque de acidificação e equalização (TAE, um reator anaeróbio de manta de lodo e fluxo ascendente (UASB, uma lagoa aerada facultativa (LAF, um equalizador de pressão e um gasômetro. O tratamento foi realizado durante 190 dias e o pH foi controlado por certos períodos de tempo, pela adição de NaOH no TAE ou no reator UASB. No reator UASB, os valores máximos e mínimos obtidos na entrada foram de 235 a 7.064 mg.L-1 para DQO; 200 a 3.913 mg.L-1 para DBO5, 500 a 11.153 mg.L-1 para STV e 4,57 a 7,75 para o pH. Na saída do reator UASB, os valores foram de 39 a 2.333 mg.L-1 para DQO; 15 a 1.300 mg.L-1 para DBO5, 272 a 2.749 mg.L-1 para STV e 6,16 a 7,93 para o pH. Os valores mínimos e máximos de vazão afluente foram de 0,18 a 1,56 L.h-1. O biogás apresentou uma produção teórica de 0,545 a de 0,602 m³.kg-1DBO5 e porcentagem de metano de 48,60 a 68,14%.It was studied the biogas production through the treatment of the wet processing coffee wastewaters (ARC in an anaerobic treatment system in laboratorial scale. The system used was composed by one acidification and equalization tank (TAE, one anaerobic upflow sludge blanket reactor (UASB, one facultative aerated pond, one equalization tank and one gas tank. The treatment was carried out for 190 days and the pH was controlled for some periods by adding NaOH inside of the TAE or in the UASB. In the UASB reactor the maximum and minimum values obtained in the inlet were 235 to 7064 mg.L-1 for COD; 200 to 3913 mg.L-1 for BOD5, 500 to 11.153 mg.L-1 for TVS and 4,57 to 7,75 for pH. In the outlet of the UASB, the values were 39 to 2333 mg.L-1 for COD; 15 to 1300 mg.L-1 for BOD5, 272 to 2749 mg.L-1 for TVS and 6,16 to 7,93 for pH. The minimum and maximum values of the inlet

  13. Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2; Medidores ultrasonicos en el flujo de agua de alimentacion para recuperar potencia termica en el reactor de la Central Nuclear Laguna Verde U1 and U2

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F. [CFE, Central Laguna Verde, Km. 42.5 Carretera Cardel-Nautla, Veracruz (Mexico)]. e-mail: francisco.tijerina@cfe.gob.mx

    2008-07-01

    The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

  14. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  15. Inventory Methods in a Conversion Plant; Methodes d'Inventaire dans un Etablissement de Transformation; Metody inventarizatsij na predpriyatii po pererabotke yadernykh materialov; Metodos de Inventario en una Planta de Transformacion

    Energy Technology Data Exchange (ETDEWEB)

    Billy, G. [Commissariat a l' Energie Atomique, Paris (France)

    1966-02-15

    inventarios es la de salvaguardar los intereses de la Comision, evaluar la cuantia de las perdidas y controlar las medidas adoptadas para el almacenaje de los desechos. Para el recuento de las existencias el inspector puede efectuar materialmente el inventario, participar en el o comprobarlo. Este ultimo procedimiento es el que da mejores resultados. Las operaciones de recuento, precedidas de una reunion preparatoria para fijar las modalidades, se efectuan en dos tiempos: la comprobacion del inventario ponderal y la comprobacion de la contabilidad fisica. Las cuestiones que pueden plantearse durante el inventario se refieren a las discrepancias en el peso y el contenido, la eleccion del equipo de pesada, los residuos y la evaluacion de las perdidas. Los metodos de inventario no deben diferir de un pais a otro. Parece preferible subrayar las dificultades que se presentan durante las operaciones y estudiar en comun los medios para evitarlas. (author) [Russian] Osnovnymi zadachami inventarizacii javljajutsja: sobljudenie interesov Komissariata; opredelenie razmerov poter'; kontrol' mer, prinimaemyh dlja hranenija othodov. Inspektirujushhee lico mozhet provodit' fakticheskuju inventarizaciju, uchastvovat' v nej ili proverjat' ee. Jetot poslednij metod daet nailuchshie rezul'taty. Operacii po uchetu, kotorym predshestvuet podgotovitel'noe soveshhanie dlja opredelenija sposobov ucheta, provodjatsjav dva jetapa: snachala proverjaetsja nalichie po vesu, azatemdoku mentacija. Voprosy, voznikajushhie v svjazi s inventarizaciej i kasajushhiesja vesa, soderzhanija poko- vok, vybora vesov dlja vzveshivanija, normy othodov i ocenki poter', razreshajutsja otpravite- lem i poluchatelem. Metody inventarizacii v odnom gosudarstve ne dolzhny sil'no otlichat'sja ot takovyh v drugom gosudarstve. Neobhodimo podcherknut' trudnosti, voznikajushhie v processe inven- tarizacii, i zhelatel'nost' otyskanija sovmestnyh putej ih preodolenija. (author)

  16. Data base formation for important components of reactor TRIGA MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

  17. Modeling and identification for the adjustable control of generation processes; Modelado e identificacion para el control autoajustable de procesos de generacion

    Energy Technology Data Exchange (ETDEWEB)

    Ricano Castillo, Juan Manuel; Palomares Gonzalez, Daniel [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1990-12-31

    The recursive technique of the method of minimum squares is employed to obtain a multivariable model of the self regressive mobile mean type, needed for the design of a multivariable, self-adjustable controller self adjustable multivariable. In this article the employed technique and the results obtained are described with the characterization of the model structure and the parametric estimation. The convergency velocity curves are observed towards the parameters` numerical values. [Espanol] La tecnica recursiva del metodo de los minimos cuadrados se emplea para obtener un modelo multivariable de tipo autorregresivo de promedio movil, necesario para el diseno de un controlador autoajustable muitivariable. En el articulo, se describe la tecnica empleada y los resultados obtenidos con la caracterizacion de la estructura del modelo y la estimacion parametrica. Se observan las curvas de la velocidad de convergencia hacia los valores numericos de los parametros.

  18. Modeling and identification for the adjustable control of generation processes; Modelado e identificacion para el control autoajustable de procesos de generacion

    Energy Technology Data Exchange (ETDEWEB)

    Ricano Castillo, Juan Manuel; Palomares Gonzalez, Daniel [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1989-12-31

    The recursive technique of the method of minimum squares is employed to obtain a multivariable model of the self regressive mobile mean type, needed for the design of a multivariable, self-adjustable controller self adjustable multivariable. In this article the employed technique and the results obtained are described with the characterization of the model structure and the parametric estimation. The convergency velocity curves are observed towards the parameters` numerical values. [Espanol] La tecnica recursiva del metodo de los minimos cuadrados se emplea para obtener un modelo multivariable de tipo autorregresivo de promedio movil, necesario para el diseno de un controlador autoajustable muitivariable. En el articulo, se describe la tecnica empleada y los resultados obtenidos con la caracterizacion de la estructura del modelo y la estimacion parametrica. Se observan las curvas de la velocidad de convergencia hacia los valores numericos de los parametros.

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  3. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  4. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  5. Simulation and tests to individual and coupled models of the reactor vessel simulator and the recirculation system for the SUN-RAH; Simulacion y pruebas a modelos individuales y acoplados del simulador de la vasija del reactor y el sistema de recirculacion para el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R.A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: rsanchez_15@yahoo.com.mx

    2004-07-01

    The present project, is continuation of the project presented in the congress SNM-2003. In this new phase of the project, they were carried out adaptive changes to the modeling and implementation of the module of the full superior of the core of the reactor, they were carried out those modeling of the generation of heat as well as of the energy transfer in the one fuel. These models present the main characteristics of the vessel of the one reactor and of the recirculation system, defined by the main phenomena that they intervene in the physical processes, in the previous version the simulation in real time it required of an extremely quick computer and without executing collateral processes. The tests are presented carried out to the different models belonging to the Simulator of the Reactor Vessel and the Recirculation system for the SUN-RAH (University Simulator of Nucleo electric with Boiling Water Reactor), as well as the results hurtled by this tests. In each section the executions of the tests and the corresponding analyses of results are shown for each pattern. Besides the above mentioned, the advantages presented by the Simulator of the reactor vessel and the recirculation system are pointed. (Author)

  6. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  8. Conocimiento y uso de metodos anticonceptivos por la poblacion femenina de una zona de salud

    Directory of Open Access Journals (Sweden)

    Maroto de Agustín Alicia

    1998-01-01

    Full Text Available FUNDAMENTO: El uso de métodos anticonceptivos está en relación, entre otros, con factores demográficos, sociales, económicos, educativos e ideológicos. El objetivo de este trabajo es conocer qué métodos anticonceptivos conocen las mujeres en edad fértil, así como la prevalencia de su uso. MÉTODOS: A partir del listado de tarjeta sanitaria se seleccionaron mediante muestreo sistemático 389 de las 5800 mujeres en edad fértil (15-45 años asignadas a un centro de salud. Previo envío de una carta, comunicando el motivo del estudio, se contactó telefónicamente con ellas para la realización de una encuesta, la cual incluía preguntas acerca del conocimiento y uso de métodos anticonceptivos, características socioculturales y actividad sexual. Las mujeres que no tenían teléfono fueron citadas en el centro de salud. RESULTADOS: Se contactó con 178 mujeres, de las que participaron 166 (tasa de respuesta de 42,7%. De ellas utilizaban algún método anticonceptivo 86 (51,8%; IC:44,2-59,4%. Sin embargo, entre mujeres con riesgo de embarazo no deseado, la prevalencia de uso era del 70,5% (IC: 62,4-78,6%, destacando el hecho de que en el grupo de mujeres de 40 a 45 años sólo utilizaran anticonceptivos el 45,4%, con una frecuencia significativamente inferior a los otros grupos de edad. Los métodos más conocidos eran el preservativo (90,4%, los contraceptivos orales (89,2% y el dispositivo intrauterino (78,3%, siendo escaso el conocimiento de otros métodos. CONCLUSIONES: La tasa de utilización de métodos anticonceptivos en mujeres con riesgo de embarazo no deseado es aceptable, si bien entre 40 y 45 años es llamativamente baja. Los métodos más conocidos son el preservativo, los contraceptivos orales y el dispositivo intrauterino.

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  10. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  12. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  14. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  15. 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.)

  16. Development of Non-Metallic Fuel Elements for a High-Temperature Gas-Cooled Reactor; Mise au point d'elements combustibles non metalliques pour un reacteur a haute temperature, refroidi par un gaz; Razrabotka nemetallicheskikh teplovydelyashchikh ehlementov dlya vysokotemperaturnogo reaktora s gazovym okhlazhdeniem; Elementos combustibles no metalicos para un reactor de temperatura elevada refrigerado por gas

    Energy Technology Data Exchange (ETDEWEB)

    Liebmann, B.; Schafer, L.; Spener, G. [NUKEM, Nuklear-Chemie und -Metallurgie G.m.b.H., Wolfgang bei Hanau, Federal Republic of Germany (Germany)

    1963-11-15

    destinados al reactor de alta temperatura refrigerado por gas de la Brown-Boveri/Krupp Reaktorbau GmbH, se investigaron y desarrollaron dos conceptos de elemento combustible. El elemento consiste en ambos casos en una esfera de grafito de 6 cm de diametro que encierra una pastilla cilindrica de combustible de unos 20 mm de diametro y 16 mm de altura. La diferencia entre ambos conceptos estriba en el tipo de combustible y en la forma de preparar las esferas de grafito. En el primero, el combustible se prepara mezclando U{sub 3}O{sub 8} y grafito, prensando esta mezcla en pastillas y haciendo que ambos componentes reaccionen en un horno al vacio a 1800{sup o}C. La razon atomica U : C es 1:45. Como este tipo de pastilla combustible no retiene cuantitativamente los productos de fision, fue necesario impregnar la esfera de grafito para hacerla impermeable y mejorar su poder de retencion. De este modo, se lograron permeabilidades del orden de 10{sup -6}cm{sup 2}/s . Con arreglo al segundo concepto, el combustible consiste en una solucion solida de UC en ZrC recubierta de una capa de ZrC. La razon molar UC : ZrC asciende a 1 : 20. La pastilla combustible se preparo del modo siguiente: se mezclaron UO{sub 2}, ZrO{sub 2} y grafito y se prensaron en pastillas que se hicieron reaccionar para obtener los carburos, que a su vez se trituraron en un molino de bolas, para volver a prensarse a 2000{sup o}C. De este modo, se alcanzaron densidades superiores al 95% del valor teorico. La memoria describe en detalle la preparacion y algunas de las propiedades fisicas de las pastillas combustibles. Se espera que este tipo de combustible retenga suficientemente los gases de fision y permita el empleo de esferas de grafito no impregnadas. La memoria examina tambien otras ventajas de esos combustibles. [Russian] V svyazi s rabotami po sovershenstvovaniyu seplovydelyayushchikh ehlementov dlya vysokotemperaturnogo reaktora s gazovym okhlazhdeniem ''Obshchestvom stroitel'stva reaktorov Braun- Boveri

  17. Reactor Radiation Loops as Large Gamma Sources; Boucles d'irradiation des reacteurs nucleaires utilisees comme sources gamma intenses; Radiatsionnye kontury yadernykh reaktorov kak moshchnye gamma-istochniki; Empleo de circuitos de irradiacion de los reactores como fuentes gamma de gran intensidad

    Energy Technology Data Exchange (ETDEWEB)

    Ryabukhina, Yu. S.

    1963-11-15

    primer lugar, se eligieron para la realizacion de esos circuitos las aleaciones de indio, metal liquido a temperatura ambiente. Se estudio el comportamiento de dos eutecticos de indio frente a algunos materiales de construccion y a principios de 1960 se construyo el primer circuito de prueba de indio-galio. Como iesultado de estudios ulteriores, se instalaron modelos de circuitos de indio-galio, con una actividad en el irradiador equivalente a unos 100 g de Ra, en el reactor IRT de la Academia de Ciencias de la Republica Socialista Sovietica de Georgia, asi'como un circuito de prueba de indio-galio-estafio in el canal del reactor IRT del Instituto de Energia Atomica de la Academia de Ciencias de la Union Sovietica. Por ultimo, en 1962, se instalo un circuito de trabajo de indio-galio-estano en el reactor IRT de la Academia de Ciencias de la Republica Socialista Sovietica de Latvia para efectuar irradiaciones en escala semiindustrial. La actividad maxima en el irradiador equivale a 30 000 g de Ra. La memoria consta de las siguientes partes: 1. ''Calculo de los circuitos de irradiacion''; en esta parte se resena la labor realizada en materia de metodos de calculo de los circuitos de irradiacion. 2. ''Modelo de un circuito de irradiacion deindio-galiodelreactor IRT de Tbilisi''; se describe el funcionamiento de este circuito. 3. ''Circuito de irradiacion de indio-galio-estano del reactor IRT de la Academia de Ciencias de la Republica Socialista Sovietica de Latvia'' ; se describe el funcionamiento de este circuito. 4. ''Perspectivas de desarrollo de los circuitos de irradiacion''; se describen los experimentos y circuitos y se presentan calculos que sugieren la posibilidad de construir circuitos de manganeso solido y circuitos con aleaciones liquidas de indio. (author) [Russian] Nachinaya s 1957 g. v SSSR provodilis' raboty po izucheniyu radiatsionnykh konturov. Byli razrabotany metody rascheta takikh sistem, izucheny vozmozhnosti razlichnykh gamma-nositelej. Vnachale byli

  18. The Non-Destructive Testing of Fuel Elements and Their Components for the United Kingdom Power-Reactor Development Programme; Controle Non Destructif des Elements Combustibles et de Leurs Parties Constitutives dans le Cadre du Programme de Developpement des Reacteurs de Puissance au Royaume-Uni; Nedestruktivnoe ispytanie teplovydelyayushchikh ehlementov i ikh komponentov dlya osushchestvleniya programmy soedinennogo korolevstva po razrabotke ehnergeticheskikh reaktorov; Ensayo No Destructivo de Elementos Combustibles y sus Componentes, en el Marco del Programa de Reactores de Potencia del Reino Unido

    Energy Technology Data Exchange (ETDEWEB)

    Mann, C. A.; Campsie, I. C. [U.K.A.E.A., Reactor Fuel Element Laboratories, Springfields, Salwick, Preston, Lancs. (United Kingdom)

    1965-10-15

    procede simple et peu couteux qui consiste a plonger la piece dans un liquide et a observer la formation de bulles. Enfin, les auteurs discutent l'emploi du krypton-85 comme radioindicateur. (author) [Spanish] Los procedimientos de ensayo que se exponen han sido establecidos en el Laboratorio de combustibles nucleares, como parte del programa del Grupo correspondiente, relativo a varillas de combustible para reactores de distintos tipos. La vainas de esas varillas consisten en tubos de acero inoxidable o aleaciones de circonio de 5 a 15 mm de diametro. a) Se describe la localizacion de fallas o grietas en los tubos. Inspeccion ultrasonica con dos sondas sumergidas. Los tubos se someten a un barrido helicoidal a gran velocidad en un tanque estacionario, con lo cual se observan y registran las senales que denotan la existencia de fallas. Para calibrar el sistema y comprobar su estabilidad, se usan como referencias unas ranuras practicadas por chisporroteo. En ciertos casos se recurre tambien a la inspeccion mediante corrientes de Foucault. Los dos metodos que se describen emplean un sistema de bobina anular de pasaje rapido y una bobina superficial con exploracion helicoidal. Para la seleccion de fases y filtrado de la senal de salida se una un circuito de puente, con frecuencias comprendidas entre 30 y 60 kHz. b) Se discute ademas la inspeccion de las dimensiones de tubos y pastillas. Se hace un estudio comparativo de diversos metodos mecanicos, neumaticos, nucleares y electronicos de medicion de las dimensiones de los tubos, y se explican las precauciones que han de adoptarse para impedir que estos se rayen. Se describen tecnicas para medir el diametro y la longitud de la circunferencia de las pastillas y se recomienda la comparacion de las circunferencias, en el caso de tubos delgados, como metodo mas ajustado a la realidad para el estudio de los problemas que plantea la existencia de huecos entre las paredes del tubo y las pastillas. El perfeccionamiento de equipo para

  19. Main activities carried out for the conversion of the reactor core TRIGA, from HEU 8.5/70 / LEU 8.5/20 to LEU 30/20; Principales actividades llevadas a cabo para la conversion del nucleo del reactor TRIGA, de HEU 8.5/70 / LEU 8.5/20 a LEU 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Flores C, J., E-mail: jorge.floresc@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In agreement with the policies of the global initiative of threats reduction (GTRI), Mexico committed that inside the reduction program of the fuel enrichment in research and test reactors (RERTR), the conversion of the core reactor TRIGA (in the nuclear centre) would be made, to use solely fuel with low enrichment ({<=} 20% U{sup 235}). To support to the execution of this commitment, a series of accords and agreements were established. The Project Agreement and Supply among the IAEA, the United States of America and Mexico was the more relevant. In this work the main activities carried out in the Instituto Nacional de Investigaciones Nucleares (ININ) with this purpose are presented. (Author)

  20. Development of a numerical code for the analysis of the linear stability of the U1 and U2 reactors of the CNLV; Desarrollo de un codigo numerico para el analisis de estabilidad lineal de los reactores de las U1 y U2 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Estrada P, C.E. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Nunez C, A.; Amador G, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Mexico D.F. (Mexico)

    2001-07-01

    The computer code ANESLI-1 developed by the CNSNS and UAM-I, has the main goal of making stability analysis of nuclear reactors of the BWR type, more specifically, the reactors of the U1 and U2 of the CNLV. However it can be used for another kind of applications. Its capacity of real time simulator, allows the prediction of operational transients, and conditions of dynamic steady states. ANESLI-1 was developed under a modular scheme, which allows to extend or/and to improve its scope. The lineal stability analysis predicts the instabilities produced by the wave density phenomenon. (Author)

  1. L’utilizzo della ricostruzione nella comunicazione del patrimonio archeologico. L’approccio, il metodo, le finalità e alcuni spunti di discussione.

    Directory of Open Access Journals (Sweden)

    Elena Bacci

    2010-05-01

    In questo contributo si focalizza l’attenzione sulla ricostruzione grafica e virtuale del patrimonio e sulle sensazioni che la ricostruzione evoca nel fruitore del messaggio culturale. La ricostruzione si attua mediante la collaborazione tra archeologo e illustratore e costituisce un momento di verifica visiva dell’interpretazione archeologica e uno strumento di comunicazione del dato archeologico fruibile a più livelli. Ciò avviene grazie allo scambio costante di informazioni (dati scientifici e proposte di ricostruzione e il confronto che ne deriva determina i metodi e le fasi di avanzamento del progetto. Il metodo si basa sull’integrazione delle immagini 3D con il disegno tradizionale ed è finalizzato alla trasposizione del dato archeologico, in modo tale da garantire alla ricostruzione il duplice requisito di soddisfazione estetica e credibilità scientifica.

  2. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  3. 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%

  4. 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).

  5. Development of computer programme for the use of empirical calculation of mining subsidence; Desarrollo informatico para utilizacion de los metodos empiricos de calculo de subsidencia minera

    Energy Technology Data Exchange (ETDEWEB)

    1999-09-01

    The fundamental objective of the project is the elaboration of a user friendly computer programme which allows to mining technicians an easy application of the empirical calculation methods of mining subsidence. As is well known these methods use, together with a suitable theoretical support, the experimental data obtained during a long period of mining activities in areas of different geological and geomechanical nature. Thus they can incorporate to the calculus the local parameters that hardly could be taken into account by using pure theoretical methods. In general, as basic calculation method, it has been followed the procedure development by the VNIMI Institute of Leningrad, a particularly suitable method for application to the most various conditions that may occur in the mining of flat or steep seams. The computer programme has been worked out on the basis of MicroStation System (5.0 version) of INTERGRAPH which allows the development of new applications related to the basic aims of the project. An important feature, of the programme that may be quoted is the easy adaptation to local conditions by adjustment of the geomechanical or mining parameters according to the values obtained from the own working experience. (Author)

  6. Analytical methods for 2,4-D (Dichlorophenoxyacetic acid) determination; Metodos analiticos para la determinacion del 2,4-D (Acido diclorofenoxiacetico)

    Energy Technology Data Exchange (ETDEWEB)

    Martinez G, M.S.M

    1999-06-01

    The 2,4-D herbicide is one of the main pesticides for controlling the bad grass in crops such as the water undergrowth. In Mexico the allowed bound of this pesticide is 0.05 mg/l in water of 2,4-D so it is required to have methods trusts and exacts, which can used in order to detected low concentration of it. In this work we show some for the conventional techniques and for establishing the 2,4-D concentrations. The UV-Vis spectrometer and liquids chromatography due that they are the most common used nowadays. Beside, we introduce a now developed technique, which is based on the neutronic activation analysis. Though use of the UV-Vis spectrometer technique it was possible target the concentrations interval between 1 and 200 mg/l. In the liquids chromatography interval was between 0.1 and 0.9, and by the neutronic activation analysis the interval was between 0.01 and 200 mg/l. (Author)

  7. An Optimised Method to Determine PAHs in a Contaminated Soil; Metodo Optimizado para la Determinacion de PAHs en un Suelo Contaminado

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

    2007-07-20

    An analytical study is presented based on an optimised method to determine selected polycyclic aromatic hydrocarbons (PAHs) by High Performance Liquid Chromatography (HPLC) with fluorescence detection. The work was focused to obtain reliable measurements of PAH in a gas work contaminated soil and was performed in the frame of the project 'Assessment of natural remediation technologies for PAHs in contaminated soils' (Spanish Plan Nacional l+D+i, CTM 2004-05832-CO2-01): First assays were focused to evaluate an initial proposed procedure by sonication extraction in the contaminated soil. Afterwards to extend the efficiency and reduce solvent and time consuming of extraction procedures, the more relevant parameters that affect the extraction step were investigated. A comparison between sonication and microwave procedures was done, and the influence of sample grinding was studied. In general, both extraction techniques led on comparable results, although sonication procedure needs to be more carefully optimised. Finally, as a final application of the optimised method, the effect of particle size on relative distribution of selected PAHs in the contaminated soil was investigated. Relative abundance of more volatile PAHs showed a decreasing according to lower grain size, while relative abundance of less volatile compounds indicated an increasing of concentration levels for lower grain size. (Author) 10 refs.

  8. The irradiation as a quarantine method for the treatment of fresh fruits; La irradiacion como metodo cuarentenario eficaz para el tratamiento de frutas frescas

    Energy Technology Data Exchange (ETDEWEB)

    Kaupert, Norma L [Comision Nacional de Energia Atomica, Ezeiza (Argentina). Dept. de Aplicaciones Tecnologicas y Agropecuarias

    1999-07-01

    The irradiation is proposed as an alternative to chemical or other physical methods for the quarantine of fresh fruit. The case of the products of the Southern part of Argentina is analysed and the economical and financial parameters for the installation and the operation of an irradiation plant are estimated. The costs are compared to those of a chemical quarantine system. (author)

  9. In-situ gamma spectrometry method for determination of environmental gamma dose; Metodo de espectrometria gamma in situ para determinacao de dose gama ambiental

    Energy Technology Data Exchange (ETDEWEB)

    Conti, Claudio de Carvalho

    1995-07-15

    This work tries to establish a methodology for germanium detectors calibration, normally used for in situ gamma ray spectrometry, for determining the environmental exposure rate in function of the energy of the incident photons. For this purpose a computer code has been developed, based on the stripping method, for the computational spectra analysis to calculate the contribution of the partial absorption of the gamma rays (Compton effect) in the active and nonactive parts of the detector. The resulting total absorption spectrum is then converted to fluence distribution in function of the energy for the photons reaching the detector, which is then used to calculate the exposure rate or kerma in air. The unfolding and fluency convention parameters are determined by detector calibration using point gamma sources. The method is validated by comparison of the results against the calculated exposure rate at a point of interest for the standards. This method is used for the direct measurement of the exposure rate distribution in function of the energy at the site, in situ measurement technic, leading to rapid results during an emergency situation and also used for indoor measurements. (author)

  10. Implementation of the Bee Colony Optimization method for the design of fuel cells; Implementacion del metodo Bee Colony Optimization para el diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J.; Ortiz S, J. J., E-mail: jaime.esquivel@fi.uaemex.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    The present work shows the results obtained after applying the Bee Colony Optimization algorithm in the design of fuel cells for a BWR. The algorithm that is implemented, works following the behavior that have the bees when pollinating a flowers field. The bees carry out an exhaustive analysis in the cell, so they leave generating diverse configurations where different fuel bars are placed with different uranium enrichments to reach a value mean, with a specific number of gadolinium bars. The behavior of the generated cell is evaluated by means of the use of the commercial code CASMO-4, which shows the variables that allow fixing if the cell fulfills the requirements. Such variables are the local potential peak factor and the neutrons multiplication factor in an infinite medium. (Author)

  11. Application of spectroscopic methods to the study of ionizing radiation effects in polymers; Aplicacion de metodos espectroscopicos para estudiar efectos de la radiacion en polimeros

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, G

    1996-12-31

    In general the interaction of ionizing radiation with polymers generates physic-chemical changes. Aiming to quantity these changes, three spectroscopic analytical techniques were used (UV, IR and EPR) and the chemical corrosion technique was used for three DSTN (CR39, Lexan and Makrofol) which were exposed to two radiation types: electrons and gammas. The effects of radiation are compared. Also a correlation between the UV and Vg results in function of dose is presented. The possible causes of the increase in chemical corrosion are discussed. (Author).

  12. Sequential method for the assessment of innovations in computer assisted industrial processes; Metodo secuencial para evaluacion de innovaciones en procesos industriales asistido por computadora

    Energy Technology Data Exchange (ETDEWEB)

    Suarez Antola, R [Universidad Catolica del Uruguay, Montevideo (Uruguay); Artucio, G [Ministerio de Industria Energia y Mineria. Direccion Nacional de Tecnologia Nuclear, Montevideo (Uruguay)

    1995-08-01

    A sequential method for the assessment of innovations in industrial processes is proposed, using suitable combinations of mathematical modelling and numerical simulation of dynamics. Some advantages and limitations of the proposed method are discussed. tabs.

  13. Implementation of sum-peak method for standardization of positron emission radionuclides; Implementacao do metodo pico-soma para padronizacao de radionuclideos emissores de positrons

    Energy Technology Data Exchange (ETDEWEB)

    Fragoso, Maria da Conceicao de Farias; Oliveira, Mercia Liane de; Lima, Fernando Roberto de Andrade, E-mail: mcfragoso@cnen.gov.br [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2015-07-01

    Positron Emission Tomography (PET) is being increasingly recognized as an important quantitative imaging tool for diagnosis and assessing response to therapy. As correct dose administration plays a crucial part in nuclear medicine, it is important that the instruments used to assay the activity of the short-lived radionuclides are calibrated accurately, with traceability to the national or international standards. The sum-peak method has been widely used for radionuclide standardization. The purpose of this study was to implement the methodology for standardization of PET radiopharmaceuticals at the Regional Center for Nuclear Sciences of the Northeast (CRCN-NE). (author)

  14. Method for the development of emergency response preparedness for nuclear or radiological accidents; Metodo para el desarrollo de la preparacion de la respuesta a emergencias nucleares o radiologicas

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    This report supplements IAEA emergency preparedness guidance published in the 1980s, and is consistent with the new international guidance. It provides practical advice for the development of an emergency response capability based on the potential nature and magnitude of the risk. In order to apply this method, emergency planners should have a good understanding of the basic radiological emergency response principles. Therefore, other applicable international guidance should be reviewed before using this report. This report provides a practical step-by-step method for developing integrated user, local and national emergency response capabilities. It can also be used as the basis for conducting an audit of an existing emergency response capability.

  15. ICP-MS: suitable method to study the metals distribution in estuarine regions; ICP-MS metodo adequado para o estudo da distribuicao de metais em regioes estuarinas

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Danilo C.; Oliveira, Arno H. [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: danilochagas@yahoo.com.br; heeren@nuclear.ufmg.br; Santos, Silvio J. dos; Brito, Veronica F.O.; Severo, Maria Isabel G. [Universidade Estadual de Santa Cruz, Ilheus, BA (Brazil). Dept. de Ciencias Biologicas

    2005-07-01

    Anthropogenic inputs of pollutants such as heavy metals into the marine environment have increased their levels to large extents within past a few decades. These pollutants tend to accumulate in the bottom sediments. As a result, ecosystems such as seaports or other industrialized coastal areas that have chronic inputs of metals have highly contaminated sediments. This characteristic has led to concerns over the ecological effects that may be associated with sediment quality. Of particular concern are toxic effects and the potential for bioaccumulation of metals in biota exposed to the sediments. The bivalves Crassostrea rhizophorae, Lucina pectinata and Mytella falcata have been used as biomonitors of trace metal contamination in two estuaries from Ilheus city, Bahia state, in Brazil. Bivalves, sediment and water samples were collected in March 2004 in Acuipe and Rio do Engenho mangroves. The proposed technique to analyze the studied matrices was the inductively coupled plasma mass spectrometry (ICP-MS). The results suggested that the studied molluscs are bioaccumulators of metals and showed the Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) as an adequate technique to determine a large range of inorganic elements, because its high sensibility and low detection limits. (author)

  16. Sub-coulomb transfer method of a nucleon for measure orbital radii; Metodo de transferencia sub-coulombiana de un nucleon para medir radios orbitales

    Energy Technology Data Exchange (ETDEWEB)

    Aguilera R, E.F.; Murillo, G.; Ramirez, J.; Avila, O. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1986-04-15

    The neutron transfer method is revised to measure neutron orbital radii and possible interest systems to apply it are determined. Its were carried out DWBA preliminary calculations for the system {sup 209} Bi(d,t) {sup 208} Bi. (Author)

  17. Method for automatic re contouring straight adaptive radiotherapy for prostate cancer; Metodo para el recontorneo automatico del recto en radioterapia adaptativa en cancer de prostata

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Vila, B.; Garcia Vicente, F.; Aguilera, E. J.

    2011-07-01

    Outline of quickly and accurately the rectal wall is important in Image Guided Radiotherapy (IGRT in the acronym) as an organ of greatest influence in limiting the dose in the planning of radiation therapy in prostate cancer. Deformabies registration methods based on image intensity can not create a correct spatial transformation if there is no correspondence between the image and image planning session. The rectal content variation creates a non-correspondence in the image intensity becomes a major obstacle to the deformable registration based on image intensity.

  18. Method to determine the activity concentration and total activity of radioactive waste; Metodo para determinar la concentracion de actividad y actividad total de desechos radiactivos

    Energy Technology Data Exchange (ETDEWEB)

    Angeles C, A

    2001-02-15

    A characteristic system of radioactive waste is described to determine the concentration of radionuclides activity and the total activity of bundles of radioactive waste. The system this integrated by three subsystems: - Elevator of drums. - Electromechanics. - Gamma spectroscopy. In the system it is analyzed waste of issuing gamma specifically, and this designed for materials of relative low density and it analyzes materials of cylindrical recipients.

  19. "Mastery Learning" Como Metodo Psicoeducativo para Ninos con Problemas Especificos de Aprendizaje. ("Mastery Learning" as a Psychoeducational Method for Children with Specific Learning Problems.)

    Science.gov (United States)

    Coya, Liliam de Barbosa; Perez-Coffie, Jorge

    1982-01-01

    "Mastery Learning" was compared with the "conventional" method of teaching reading skills to Puerto Rican children with specific learning disabilities. The "Mastery Learning" group showed significant gains in the cognitive and affective domains. Results suggested Mastery Learning is a more effective method of teaching…

  20. El uso del metodo MICMAC y MACTOR analisis prospectivo en un area operativa para la busqueda de la excelencia operativa a traves del Lean Manufacturing

    Directory of Open Access Journals (Sweden)

    Juan Baldemar Garza Villegas

    2011-07-01

    Full Text Available Presents the MICMAC and MACTOR prospective analysis study in an operational area which looks for world-class management with reference to the concepts of lean manufacturing. The results of a qualitative analysis and his conclusions are presented.

  1. The comparison of two methods to obtain human oral keratinocytes in primary culture; Comparacao de dois metodos de obtencao celular para cultura primaria de queratinocitos bucais humanos

    Energy Technology Data Exchange (ETDEWEB)

    Klingbeil, Maria Fatima Guarizo

    2006-07-01

    The therapeutic procedures frequently used in oral treatments for the pathological diseases are surgical, resulting in failures of the mucosal continuity.The possibility to obtain transplantable oral epithelia from an in vitro cell culture opens new utilization perspectives not only to where it comes from, but also as a reconstructive material for other parts of the human body, such as: urethra, epithelia corneo-limbal, cornea, ocular surface. Many researchers still use controversial methods for obtaining cells. It was therefore evaluated and compared the efficiency in both methods: enzymatic and direct explant to obtain oral keratinocytes from human oral mucosa. Fragments of intra oral epithelial tissues from healthy human subjects, undergoing dental surgeries, were donated to the research project. The keratinocytes were cultivated over a feeder-layer from a previously irradiated 3T3 Swiss albino fibroblasts. In this study it was compared the time needed in the cell obtention, the best cell amount between both methods, the life-span, the cell capacity to form an in vitro epithelia and its morphologic structure. The results in the assessment of both methods have shown the possibility to obtain keratinocytes from a small oral fragment, but at the same time we may verify the advantages and peculiar restrictions for each one of both analyzed methods. (author)

  2. Calculation methods of reactivity using derivatives of nuclear power and Filter fir; Metodos para o calculo da reatividade usando derivadas da potencia nuclear e o filtro FIR

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, Daniel Suescun

    2007-07-01

    This work presents two new methods for the solution of the inverse point kinetics equation. The first method is based on the integration by parts of the integral of the inverse point kinetics equation, which results in a power series in terms of the nuclear power in time dependence. Applying some conditions to the nuclear power, the reactivity is represented as first and second derivatives of this nuclear power. This new calculation method for reactivity has special characteristics, amongst which the possibility of using different sampling periods, and the possibility of restarting the calculation, after its interruption associated it with a possible equipment malfunction, allowing the calculation of reactivity in a non-continuous way. Apart from this reactivity can be obtained with or without dependency on the nuclear power memory. The second method is based on the Laplace transform of the point kinetics equations, resulting in an expression equivalent to the inverse kinetics equation as a function of the power history. The reactivity can be written in terms of the summation of convolution with response to impulse, characteristic of a linear system. For its digital form the Z-transform is used, which is the discrete version of the Laplace transform. In this method it can be pointed out that the linear part is equivalent to a filter named Finite Impulse Response (Fir). The Fir filter will always be, stable and non-varying in time, and, apart from this, it can be implemented in the non-recursive way. This type of implementation does not require feedback, allowing the calculation of reactivity in a continuous way. The proposed methods were validated using signals with random noise and showing the relationship between the reactivity difference and the degree of the random noise. (author)

  3. Calibration methods of plane-parallel ionization chambers used in electron dosimetry; Metodos de calibracao de camaras de ionizacao de placas paralelas para dosimetria de feixes de eletrons

    Energy Technology Data Exchange (ETDEWEB)

    Bulla, Roseli Tadeu

    1999-07-01

    The use of linear accelerators in radiotherapy is of great importance in Medicine, and according to international recommendations the electron beam dosimetry has to be performed using plane-parallel ionization chambers, previously calibrated in standard gamma radiation fields at accredited laboratories. In this work, calibration methods of plane-parallel ionization chambers used in dosimetry procedures of high energy electron beams of clinical accelerators were presented, tested and intercompared. The experiments were carried out using gamma radiation beams of {sup 60} Co at the Calibration Laboratory of Clinical Dosemeters at IPEN and electron beams od 4 to 16 MeV at the Radiotherapy Department of Hospital Israelita Albert Einstein, Sao Paulo. A method was chosen to be established at IPEN. Proposals of the calibration procedure, calibration certificate and data sheets are presented. (author)

  4. Validation of analytical method to calculate the concentration of conjugated monoclonal antibody; Validacao de metodo analitico para calculo de concentracao de anticorpo monoclonal conjugado

    Energy Technology Data Exchange (ETDEWEB)

    Alcarde, Lais F.; Massicano, Adriana V.F.; Oliveira, Ricardo S.; Araujo, Elaine B. de, E-mail: lais_alcarde@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The objective of this study was to develop a quantitative analytical method using high performance liquid chromatography (HPLC) to determine the antibody concentration in conjunction with bifunctional chelator. Assays were performed using a high performance liquid chromatograph, and the following conditions were used: flow rate of 1 mL / min, 15 min run time, 0.2 M sodium phosphate buffer pH 7.0 as the mobile phase and column of molecular exclusion BioSep SEC S-3000 (300 x 7.8 mm, 5 μM - Phenomenex). The calibration curve was obtained with AcM diluted in 0.2 M sodium phosphate buffer pH 7.0 by serial dilution, yielding the concentrations: 400 μg/mL, 200 μg/mL, 100 μg/mL, 50 μg/mL, 25 μg/mL and 12.5 μg/mL. From the calibration curve calculated the equation of the line and with it the concentration of the immunoconjugate. To ensure the validity of the method accuracy and precision studies were conducted. The accuracy test consisted in the evaluation of 3 samples of known concentration, being this test performed with low concentrations (50 μg/mL), medium (100 μg/mL) and high (200 μg/mL). The precision test consisted of 3 consecutive measurements of one sample of known concentration, subject to the conditions set forth above for the other tests. The correlation coefficient of the standard curve was greater than 97%, the accuracy was satisfactory at low concentrations as well as accuracy. The method was validated by showing it for the accurate and precise determination of the concentration of the immunoconjugate. Furthermore, this assay was found to be extremely important, because using the correct mass of the protein, the radiochemical purity of the radioimmunoconjugate was above 95% in all studies.

  5. The Muhlbauer method for pipeline risk management in onshore environment; O metodo de Muhlbauer para gerenciamento de risco em linhas de dutos em ambiente 'onshore'

    Energy Technology Data Exchange (ETDEWEB)

    Schafer, Alexandro G.; Miguelis, Paula M.F. [UNIPAMPA, RS (Brazil)

    2008-07-01

    There are several methods for the risk assessment and risk management applied to pipelines, among them the Muhlbauer's Method. W. Kent Muhlbauer is an internationally recognized authority on pipeline risk management. He made a detailed identification about 300 distinct conditions that influence the risk assessment in pipelines and he proposed a score system that is known as method of Muhlbauer. The purpose of this model is to evaluate the public exposure to the risk and identify ways for management that risk in fact. The assessment is made by the attribution of quantitative values to the several items that influences in the pipeline risk. This paper approaches the Muhlbauer's basic model for risk assessment and management in pipelines. In the beginning, the basic model for risk assessment is presented, and methodology for pipelines in onshore environment is detailed. After, presents major items in risk assessment and this relative score. Finally, present the additional modules for Muhlbauer's method customizing. (author)

  6. Computer simulating for oil fields with artificial elevation method by electrical submersible pump; Simulacao computacional para pocos de petroleo com metodo de elevacao artificial por bombeio centrifugo submerso

    Energy Technology Data Exchange (ETDEWEB)

    Batista, Evellyne da Silva; Maitelli, Andre Laurindo [Universidade Federal do Rio Grande do Norte (UFRN), Natal, RN (Brazil); Costa, Rutacio de Oliveira [PETROBRAS S.A., Natal, RN (Brazil). Unidade de Negocio RN/CE; Barbosa, Tiago de Souza [Universidade Federal do Rio Grande do Norte (UFRN), Natal, RN (Brazil). Dept. de Engenharia de Computacao

    2008-07-01

    The architecture heterogeneities and petrophysical properties of carbonate reservoirs result from a combination of platform morphology, related depositional environments, relative sea level changes and diagenetic events. The reservoir layering built for static and dynamic modelling purposes should reflect the key heterogeneities (depositional or diagenetic) which govern the fluid flow patterns. The layering needs to be adapted to the goal of the modelling, ranging from full field computations of hydrocarbon volumes, to sector-based fine-scale simulations to test the recovery improvement. This paper illustrates various reservoir layering types, including schemes dominated by depositional architecture, and those more driven by the diagenetic overprint. The examples include carbonate platform reservoirs from different stratigraphic settings (Tertiary, Cretaceous, Jurassic and Permian) and different regions (Europe, Africa and Middle East areas). This review shows how significant stratigraphic surfaces (such as sequence boundaries or maximum flooding) with their associated facies shifts, can be often considered as key markers to constrain the reservoir layering. Conversely, how diagenesis (dolomitization and karst development), resulting in units with particular poroperm characteristics, may significantly overprint the primary reservoir architecture by generating flow units which cross-cut depositional sequences. To demonstrate how diagenetic processes can create reservoir bodies with geometries that cross-cut the depositional fabric, different types of dolomitization and karst development are illustrated. (author)

  7. Method of compensation spires for the detection of the diamagnetic effect in a Tokamak; Metodo de espiras de compensacion para la deteccion del efecto diamagnetico en un Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Colunga S, S

    1990-09-15

    In this report the classical detection method of the diamagnetic effect by means of a rolled spire on the discharges chamber in the poloidal direction and the difficulties related with this are analyzed. An alternative method that increases considerably the detection sensibility of the diamagnetic effect and that for its simplicity it is quite attractive for its application to the Tokamak Novillo of the ININ is presented. (Author)

  8. Development of method to chemical separation of gallium-67 by thermal diffusion technique; Desenvolvimento de metodo para separacao quimica de galio-67 pela tecnica de difusao termica

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Patricia de Andrade

    2012-07-01

    Radioisotopes of gallium have been studied and evaluated for medical applications since 1949. Over the past 50 years {sup 67}Ga has been widely used in the diagnosis of various diseases, including acute and chronic inflammatory lesions, bacterial or sterile and several types of tumors. In Brazil 30% of clinics that provide services for Nuclear Medicine use {sup 67}Ga citrate and the demand for 67{sup G}a at IPEN-CNEN/SP is 37 GBq (1 Ci)/week. The {sup 67}Ga presents physical half-life of 3.26 days (78 hours) and decays 100% by electron capture to stable {sup 67}Zn. Its decay includes the emission of {gamma} rays with energies of 93.3 keV (37%), 184.6 keV (20.4%), 300.2 keV (16.6%) and 888 keV (26%). In the past {sup 67}Ga was produced by the reaction {sup 68}Zn (p, 2n) {sup 67}Ga at IPEN-CNEN/SP. After irradiation, the target was dissolved in concentrated HCl and the solution percolated through a cationic resin DOWEX 50W-X8, 200-400 mesh, conditioned with 10 mol L{sup -1} HCl. Zinc, nickel and copper were eluted in 10 mol L{sup -1} HCl and {sup 67}Ga 3.5 mol L{sup -1} HCl. The final product was obtained as {sup 67}Ga citrate. This work presents a new, fast, direct and efficient method for the chemical separation of 67{sup G}a by thermal diffusion (heating of the target) combined with concentrated acetic acid extraction. Purification was performed by ion exchange chromatography. Natural zinc electrodeposition was performed on nickel/copper plates as substrate and the zinc deposits were adherent to the substrate, slightly shiny and uniform. The targets were irradiated with 26 MeV protons and integrated current of 10 {mu}A.h. After irradiation, the targets were heated at 300 deg C for 2 hours and placed in contact with concentrated acetic acid for 1 hour. The average yield of extraction of {sup 67}Ga was (72 {+-} 10)%. This solution was evaporated and the residue was taken up in 0.5 mol L{sup -1} NH{sub 4}OH. The 67{sup G}a was purified on cationic resin Dowex 50WX8 in NH{sub 4}OH medium. The {sup 67}Ga recovery was (98 {+-} 2)%. This solution was evaporated and taken up in 0.1 mol L{sup -1} HCl. The chemical purity was evaluated by ICP-OES that resulted in (2 {+-} 1) {mu}g mL{sup -1} of zinc. The concentration of iron, copper and nickel was lower than the detection limits and also than the utilization limits for {sup 67}Ga. The radionuclidic purity was greater than (99.9%). This method showed to be suitable to obtain high purity {sup 67}Ga in less aggressive chemical conditions than before. (author)

  9. An experimental-numerical method for comparative analysis of joint prosthesis; Un metodo numerico-experimental para el analisis comparativo de protesis articulares

    Energy Technology Data Exchange (ETDEWEB)

    Claramunt, R.; Rincon, E.; Zubizarreta, V.; Ros, A.

    2001-07-01

    The difficulty that exists in the analysis of mechanical stresses in bones is high due to its complex mechanical and morphological characteristics. This complexity makes generalists modelling and conclusions derived from prototype tests very questionable. In this article a relatively simple comparative analysis systematic method that allow us to establish some behaviour differences in different kind of prosthesis is presented. The method, applicable in principle to any joint problem, is based on analysing perturbations produced in natural stress states of a bone after insertion of a joint prosthesis and combines numerical analysis using a 3-D finite element model and experimental studies based on photoelastic coating and electric extensometry. The experimental method is applied to compare two total hip prosthesis cement-free femoral stems of different philosophy. One anatomic of new generation, being of oblique setting over cancellous bone and the other madreporique of trochantero-diaphyseal support over cortical bone. (Author) 4 refs.

  10. Method for evaluating the applicability and application rate wastes in soil; Metodo para avaliacao da aplicabilidade e taxa de aplicacao de residuos em solo

    Energy Technology Data Exchange (ETDEWEB)

    Linhares, Monica Moreira [Partime (Brazil); Seabra, Paulo Negrais [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas

    1991-01-01

    Land treatment of refinery solid wastes has been practiced in the last few years by some PETROBRAS refineries. The biodegradation process is dynamic and complex, and incorrect monitoring may destroy the potential of the soil. Due to great differences both in terms of soils and of residues, each case must be monitored individually. We therefore developed a monitoring method for land farming systems based on the impact of the waste on indigenous soil microbial populations. The method is extremely simple, requiring only a gas chromatograph to evaluate the CO{sub 2} evolution, and can detect possible causes for low efficiency biodegradation processes. Conditions should then be altered, to ensure adequate functioning of the system.The method also allows for determination of the applicability of candidate wastes and of acceptable waste application rates in soil. (author) 2 refs., 7 figs.

  11. Optimization of a radiochemistry method for plutonium determination in biological samples; Optimizacion del metodo radioquimico para determinar plutonio en muestras biologicas

    Energy Technology Data Exchange (ETDEWEB)

    Cerchetti, Maria L; Arguelles, Maria G [Comision Nacional de Energia Atomica, Ezeiza (Argentina). Laboratorio de Dosimetria Personal y de Area

    2005-07-01

    Plutonium has been widely used for civilian an military activities. Nevertheless, the methods to control work exposition have not evolved in the same way, remaining as one of the major challengers for the radiological protection practice. Due to the low acceptable incorporation limit, the usual determination is based on indirect methods in urine samples. Our main objective was to optimize a technique used to monitor internal contamination of workers exposed to Plutonium isotopes. Different parameters were modified and their influence on the three steps of the method was evaluated. Those which gave the highest yield and feasibility were selected. The method involves: 1-) Sample concentration (coprecipitation); 2-) Plutonium purification; and 3-) Source preparation by electrodeposition. On the coprecipitation phase, changes on temperature and concentration of the carrier were evaluated. On the ion-exchange separation, changes on the type of the resin, elution solution for hydroxylamine (concentration and volume), length and column recycle were evaluated. Finally, on the electrodeposition phase, we modified the following: electrolytic solution, pH and time. Measures were made by liquid scintillation counting and alpha spectrometry (PIPS). We obtained the following yields: 88% for coprecipitation (at 60 C degree with 2 ml of CaHPO{sub 4}), 71% for ion-exchange (resins AG 1x8 Cl{sup -} 100-200 mesh, hydroxylamine 0.1N in HCl 0.2N as eluent, column between 4.5 and 8 cm), and 93% for electrodeposition (H{sub 2}SO{sub 4}-NH{sub 4}OH, 100 minutes and pH from 2 to 2.8). The expand uncertainty was 30% (NC 95%), the decision threshold (Lc) was 0.102 Bq/L and the minimum detectable activity was 0.218 Bq/L of urine. We obtained an optimized method to screen workers exposed to Plutonium. (author)

  12. Utilizacion del metodo de hidrolisis de los amilostatolitos para el diagnostico de resistencia al calor, la sequia y la salinidad en el tomate

    Directory of Open Access Journals (Sweden)

    Jorge A. Sánchez

    2000-01-01

    Full Text Available Se propone 5 indices de calidad que penni- ten estimar certerarnentel nivel de calidad y po- tencial productivo de una plantaci6n forestal. £1 primer indice es el de calidad general y debe uti- lizarse preferiblemente en plantaciones no ralea- das. £ste indiceproducira valores de 1.0-4.0. Valo- res cercanos a 1 serlin de plantaciones de la mas al- ta calidad. £1 segundo es el de calidad de cosecha y se basa en la cantidad de individuos presentes por ha de calidad 1 y 2. Una plantaci6n con 400 indi- viduos de calidad 1 +2 seria excelente y con menos de 200 individuos seria de mala calidad. £1 tercero es el de calidad maxima, que busca reflejar en que proporci6n la plantaci6n se aproxima al maximo numero posible de individuos de calidad 1 y de 40 cm de dap que puede contener una plantaci6n fo- festal ala cosecha; es basado en N 1=250 Y genera harvalores porcentuales, donde una plantaci6n con >90% seria excelente y 1600 trozas/ha de calidad 1+2 son excelentes y aquellas <800 trozas son inaceptables. Por ultimo se incluye el indice de calidad de productividad, que permite valorar la calidad de una plantaci6n en

  13. Planetary method to measure the neutrons spectrum in lineal accelerators of medical use; Metodo planetario para medir el espectro de neutrones en aceleradores lineales de uso medico

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Benites R, J. L., E-mail: fermineutron@yahoo.com [Centro Estatal de Cancerologia de Nayarit, Servicio de Seguridad Radiologica, Calzada de la Cruz 118 Sur, 63000 Tepic, Nayarit (Mexico)

    2014-08-15

    A novel procedure to measure the neutrons spectrum originated in a lineal accelerator of medical use has been developed. The method uses a passive spectrometer of Bonner spheres. The main advantage of the method is that only requires of a single shot of the accelerator. When this is used around a lineal accelerator is necessary to operate it under the same conditions so many times like the spheres that contain the spectrometer, activity that consumes enough time. The developed procedure consists on situating all the spheres of the spectrometer at the same time and to realize the reading making a single shot. With this method the photo neutrons spectrum produced by a lineal accelerator Varian ix of 15 MV to 100 cm of the isocenter was determined, with the spectrum is determined the total flow and the ambient dose equivalent. (Author)

  14. Automatic method detection of artifacts for control of tomographic uniformity on SPECT; Metodo automatico de dteccion de artefactos para el control de la uniformidad tomografica en SPECT

    Energy Technology Data Exchange (ETDEWEB)

    Reynes Llompart, G.; Puchal, R.

    2013-07-01

    The objective of this work is the find an automatic method for the detection and classification of artifacts produced in tomographic uniformity, extracting the characteristics necessary to apply a classification algorithm using pattern recognition techniques. The method has been trained and validated with synthetic images and tested with real images. (Author)

  15. Torpedo Base - a new method for installation of a conductor casing; Base Torpedo - um novo metodo para instalacao do revestimento condutor

    Energy Technology Data Exchange (ETDEWEB)

    Nogueira, Emmanuel Franco; Borges, Alexandre Thomaz [PETROBRAS. E and P Engenharia de Producao. Gerencia de Perfuracao e Operacoes Especiais (Brazil)], e-mails: efranco@petrobras.com.br, atborges@petrobras.com.br; Medeiros Junior, Cipriano Jose de [PETROBRAS. Engenharia. Gerencia de Geociencias (Brazil)], e-mail: cipri.intec@petrobras.com.br; Machado, Rogerio Diniz; Souza, Ebenezer Viana de [PETROBRAS. E and P - SERV. Gerencia de Ancoragem (Brazil)], e-mail: rogeriodm@petrobras.com.br, ebnezer_souza@petrobras.com.br

    2008-12-15

    In the last years, PETROBRAS has been optimizing and reducing the cost of the mooring systems through the use of torpedo anchors. It consists of a tubular pile with heavy weight ballast, installed by a process that uses the energy generated by the free fall of the pile from a supply vessel. It is a low cost anchoring concept, with simple fabrication and installation. More than two hundred torpedo anchors were installed offshore Brazil in the last four years. Based on the success of the torpedo anchors for mooring purposes PETROBRAS Exploration and Production department decided to employ the same idea for installing the conductor casing. The benefit was that this installation could be accomplished from an anchor-handling vessel saving important rig time. Another advantage with this installation method is a better foundation provided since no soil is removed during this process. The paper will provide an overview of PETROBRAS Torpedo Base experience in the Albacora Leste field as well as a few other applications of this technology for the production side. After the final installation of two prototypes in the Albacora Leste Field, in Campos Basin, at the end of 2004, the Torpedo Base project was approved by the clients and by PETROBRAS's technological area, having its acquisition requested by five assets (management areas responsible for oil fields located in that basin), in a total of 18 Torpedo Base. The difficulties related to the hiring of services and logistics faced in order to make the installation of the equipment feasible made the E and P technological area to choose to monitor the acquisition and installation of these 18 bases, so as to ensure a smooth transition between the project authors and the technicians in charge of operation of the equipment, ensuring, therefore, the consolidation of the technology. This paper reports PETROBRAS's experience with Torpedo Bases during the installation of these 18 bases in several fields in Campos Basin, throughout almost two years. There shall be presented the problems faced, solutions for them and improvements implemented aiming at optimizing not only the equipment but also its installation process. (author)

  16. Electret ionization chamber: a new method for detection and dosimetry of thermal neutrons; Camara de ionizacao de eletretos: um novo metodo para deteccao e dosimetria de neutrons termicos

    Energy Technology Data Exchange (ETDEWEB)

    Ghilardi, A J.P.

    1988-12-31

    An electret ionization chamber with boron coated walls is presented as a new method for detecting thermal neutrons. The efficiency of electret ionization chambers with different wall materials for the external electrode was inferred from the results. Detection of slow neutrons with discrimination against the detection of {gamma}-rays and energetic neutrons was shown to depend on the selection of these materials. The charge stability over a long period of time and the charge decay owing to natural radiation were also studied. Numerical analysis was developed by the use of a micro-computer PC-XT. Both the experimental and numerical results show that the sensitivity of the electret ionization chamber for detection of thermal neutrons is comparable with that of the BF{sub 3} ionization chamber and that new technologies for deposition of the boron layer will produce higher efficiency detectors. (author). 102 refs, 32 fig, 10 tabs.

  17. Validation of a method to determine methylmercury in fish tissues using gas chromatography; Validacion de un metodo para determinar metilmercurio en tejido de pescado por cromatografia de gases

    Energy Technology Data Exchange (ETDEWEB)

    Vega Bolannos, Luisa O; Arias Verdes, Jose A; Beltran Llerandi, Gilberto; Castro Diaz, Odalys; Moreno Tellez, Olga L [Instituto de Nutricion e Higiene de los Alimentos, La Habana (Cuba)

    2000-07-01

    We validated a method to determine methylmercury in fish tissues using gas chromatography with an electron capture detector as described by the Association of Official Analytical Chemist (AOAC) International. The linear curve range was 0.02 to 1 g/ml and linear correlation coefficient was 0.9979. A 1 mg/kg methylmercury-contaminated fish sample was analyzed 20 times to determine repeatability of the method. The quantification limit was 0.16 mg/kg and detection limit was 0.06 ppm. Fish samples contaminated with 0.2 to 10 mg/kg methylmercury showed recovery indexes from 94.66 to 108.8%.

  18. 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

  19. 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

  20. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo a baja presion (LPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Membrillo G, O. E.; Chavez M, C., E-mail: garzo1012@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present work presents the modeling and simulation of the aspersion system to low pressure of reactor of the nuclear power plant of Laguna Verde using the nuclear code RELAP/SCDAP. The objective of the emergency systems inside a nuclear reactor is the cooling of the core, nor caring the performance of any other emergency system in the case of an accident design base for coolant loss. To obtain a simulation of the system is necessary to have a model based on their main components, pipes, pumps, valves, etc. This article describes the model for the simulation of the main line and the test line for the HPCS. At the moment we have the simulation of the reactor vessel and their systems associated to the nuclear power plant of Laguna Verde, this work will allow to associate the emergency system model LPCS to the vessel model. The simulation of the vessel and the emergency systems will allow knowing the behavior of the reactor in the stage of the coolant loos, giving the possibility to analyze diverse scenarios. The general model will provide an auxiliary tool for the training in classroom and at distance in the operation of nuclear power plants. (Author)

  1. Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico; Evaluacion de la aptitud para el servicio de la piscina del reactor TRIGA Mark III del Instituto Nacional de Investigaciones Nucleares de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J. [Departamento de Sintesis y Caracterizacion de Materiales del ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

  2. Preparation of mandatory documentation before the start up of the RA-0 `zero power` nuclear reactor at Cordoba National University; Preparacion de la documentacion mandatoria para la puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Keil, W M; Pezzi, N

    1992-12-31

    Before the start up of the RA-0 `zero power` nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the `70, a work program for the future operational training personnel was elaborated. Based on the Authority`s applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author). [Espanol] Con motivo de la nueva puesta en servicio del REACTOR NUCLEAR RA-0 fue necesario elaborar la documentacion mandatoria requerida por la Autoridad Regulatoria Nacional. Siguiendo los lineamientos de las normas y recomendaciones vigentes e incluyendo criterios propios en lo que debia ser el contenido final de dicha documentacion, fue preparado lo que se ha denominado el INFORME DE SEGURIDAD DEL REACTOR NUCLEAR RA-0. Este documento que se describe en este trabajo, si bien contiene las habituales descripciones de todos los Informes de Seguridad, incluye otros aspectos que no siendo requeridos expresamente en el mismo, han dado una mayor coherencia a la conformacion de todos los aspectos que interrelacionan las areas de seguridad fisica, radiologica, nuclear y de control de materiales nucleares bajo salvaguardias. (Autor).

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  5. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model; Aplicacao da teoria de perturbacao para calculos de sensibilidade em nucleos de reatores PWR, usando um modelo de dois canais

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, A.C.J.G. de

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs.

  6. Development and implementation of a new pneumatic transfer system for materials irradiation at IEA-R1 reactor; Desenvolvimento e implementacao de um novo sistema pneumatico de transferencia para irradiacao de materiais no reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Fernando, Alberto de Jesus

    2011-07-01

    Pneumatic Transfer Systems (PTS) are classified as mechanical equipment largely operated all over the world for transport of a huge sort of objects, samples and materials located at nearly terminals or even at separated ones. System applicability is often recognized in many activities, such as medicine (hospital settings, clinical analysis labs), industry (steel, automobiles, mining, chemical, food, construction), trading (gas station, movies, supermarkets, banks, e-commerce) and federal agencies (post services, federal courts, public enterprises). In the nuclear settings, PTS shows also a vast array of applications, being a part of radioisotope production, as well as short-lived radiopharmaceuticals, including 67 Ga, 201 Tl, 18 F and 123 I-ultra pure. Besides, PTS are also used at radioactive waste management plants and research institutes that apply neutron activation analysis (NAA). This work was directed toward the design and operation of a new PTS for the IEA-R1 nuclear research reactor settled at Instituto de Pesquisas Energeticas e Nucleares (IPEN) for NAA application. With this aim, it was calculated the charge of reactor core grid plate and sample transport testing. Neutron flux at irradiating position was determined as 3,70 {+-} 0,26 10{sup 12} n cm{sup -2} s{sup -1}. (author)

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  8. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  9. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  11. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  12. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  13. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  14. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  15. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  16. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  17. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  18. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  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. The Application of Various Nondestructive Testing Methods to Fuel Elements of the Orgel Type; Application des Differentes Methodes d'Essais Non Destructifs aux Elements Combustibles du Type Orgel; Primenenie razlichnykh nedestruktivnykh metodov ispytanij k toplivnym ehlementam tipa ''orgel''; Aplicacion de Distintos Metodos de Ensayo No Destructivo a los Elementos Combustibles de Tipo Orgel

    Energy Technology Data Exchange (ETDEWEB)

    Bonnet, P.; Jansen, J. [EURATOM, C.C.R., Ispra (Italy)

    1965-09-15

    methode a un capteur montre quelles sont les limites d'utilisation de ces deux methodes. Defauts transversaux; La methode a un capteur appliquee a cette recherche fait l'objet d'une breve description. Entrainement mecanique: Le memoire fait etat d'un banc mecanique, type laboratoire, pour la recherche des criteres d'essais et d'un banc de passage des tubes, de type semi-industriel, pour l'examen en continu avec enregistrement des defauts. Les difficultes rencontrees et les moyens misenceuvrepoury remediersont evoques. c) Tests par radiographie. Cette methode fera l'objet d'un memoire special; seuls sont indiques les resultats obtenus sur tubes de force ou sur tubes de gaines. d) Test divers Les test finaux sur elements combustibles complets peuvent se resumer en deux parties. Tests d'etancheite au moyen du test a l'helium developpe par la SOGEV. Un dispositif permet d'examiner simultanement 4 ou 6 elements combustibles ou separement 2 ou 3. Tests classiques de radiographie; l'accent est mis plus particulierement sur l'etude des soudures; on donne la marge des conditions experimentales pour obtenir une bonne definition. En conclusion, les auteurs presentent un projet de chaine de controle semi-industrielle avec les differentes possibilites de traitements et essaient de degager une certaine philosophie des tests non destructifs appliques aux elements combustibles. (author) [Spanish] La memoria presenta los distintos metodos empleados para deteccion de fallas (de tipo dimensional o estructural) en vainas de elementos combustibles. Los autores describen, asimismo, los ensayos finales con elementos combustibles completos, en particular radiografia de las soldaduras y pruebas de estanqueidad. Este tema se ha tratado ya en forma parcial. El estudio de las caracteristicas dimensionales de las vainas lisas de SAP (producto de aluminio sinterizado) ha sido objeto de trabajos bastante avanzados. La memoria examina en particular los siguientes temas: 1. Medicion de diametros internos

  1. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  2. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  3. Development of computational program for studying the reactor control system in PWR plants; Desenvolvimento de um programa computacional para estudo do sistema de controle do reator em plantas PWR

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Jose Ricardo de; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    In this work a computational program is presented which has been developed for specific application on the study of the reactor control system of a typical PWR plant. As to the basic function of simulating power transients the program has the following structure: a representative mathematical model of the dynamic and stationary behaviors of the primary circuit; a group of equations associated to the reactor power control and system pressure control; screens for the entry of reference data as well as of control blocks and control bar speed programming module parameters; main entering screens for the configuration of the excitement/transient function as well as of simulation time and control mood; and graphical output of all the process variables incorporated to the model. As premise it has been considered as sufficient the modeling of the primary circuit, a differential equation being used which associates the average temperature of the coolant within the steam generator with the potency transferred to the secondary circuit, denominated 'secondary potency', as an interface with the secondary circuit. Every transient - ramp or step - is established upon the 'turbine power' variable, which in turn is related to the 'secondary power' variable by means of a differential equation that represents a first - order delay, having adjustable parameters on the data - entry screen. In the neutronic model as defined for the reactor, the reactivity feedback effects due to primary circuit pressure variation, as well as fuel and coolant temperature variation, were taken into consideration. Thermo-hydraulics constants and project data taken from the available bibliography, adapted to a particular small PWR unit conception , were employed for loading the program. With the open-loop simulation results a positive qualitative evaluation of the program was obtained, in comparison to published results related to simulators bearing equal purposes, more

  4. 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.)

  5. Spectrographic Determination of Impurities in Ceramic Materials for Nuclear Fusion Reactors. I. Analysis of Alumina; Determinacion Espectrografica de impurezas en materiales ceramicos para fusion nuclear. I.- Analisis de alumina

    Energy Technology Data Exchange (ETDEWEB)

    Rucandio, M I; Roca, M; Melon, A

    1990-07-01

    The determination of minor and trace elements in the aluminium oxide considered as possible ceramic material in thermonuclear fusion reactors has been studied. The concentration ranges are 0.1 - 0.3 * for Ca, Si and Y, and at the ppm level for Co, Cr, Fe, Hf, K, Li, Mg, Mn, Na, Ni, Se, Ta, Ti, V and Zr. Atomic emission spectroscopy with direct current ore excitation and photographic detection has been employed. For Hf, Mg, Ta, Ti, V and Zr the use of 40% of copper fluoride as a carrier and of Nb as lnternal standard provide suitable sensitivities and precessions, while for the rest of elements the bent results are obtained with graphite powder in different proportions and Rb or Sn as internal standard. (Author) 7 refs.

  6. Solution of the neutron diffusion equation to study the 3D distribution of power, applied to nuclear reactors; Solucao da equacao de difusao de neutrons para o estudo da distribuicao de potencia em 3D, aplicado a reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Danilo Leite

    2013-07-01

    This work aims to present a study about the power distribution behavior in a PWR type reactor, considering both intensity and migration of power peaks due to insertion of control rods into the core. Employing the multidimensional steady-state neutron diffusion equation in order to simulate the neutron flux, and using the Finite Difference Method. Furthermore, based on the axial power distribution on the largest heat flux rod, is carried out thermal analysis of this rod and associated coolant channel. For this purpose is employed the FueLRod{sub 3}D code, it uses the Finite Element Method to model the fuel rod and the associated coolant channel, allowing the thermohydraulics simulation of a single rod discretized in three dimensions, considering the heat flux from the pellet, crossing the gap and the cladding until it reaches the coolant. (author)

  7. Flow visualization system for wind turbines without blades applied to micro reactors; Sistema de visualização de escoamento para turbinas sem lâminas aplicada a microrreatores

    Energy Technology Data Exchange (ETDEWEB)

    Santos, G.S.B., E-mail: siqueira.gsb@gmail.com [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), São José dos Campos, SP (Brazil); Guimarães, L.N.F. [Instituto de Estudos Avançados (IEAv), São José dos Campos, SP (Brazil). Departamento de Ciência e Tecnologia Aeroespacial; Placco, G.M. [Instituto Tecnológico de Aeronáutica (PG/CTE/ITA), São José dos Campos, SP (Brazil). Departamento de Ciência e Tecnologia Aeroespacial

    2017-07-01

    Flow visualization systems is a tool used in science and industry for characterization of projects that operate with drainage. This work presents the design and construction of a flow visualization system for passive turbines used in advanced fast micro reactors. In the system were generated images where it is possible to see the supersonic and transonic flow through the turbine disks. A test bench was assembled to generate images of the interior of the turbine where the flow is supersonic, allowing the study of the behavior of the boundary layer between disks. It is necessary to characterize the boundary layer of this type of turbine because its operation occurs in the transfer of kinetic energy between the fluid and the disks. The images generated, as well as their analyzes are presented as a result of this work.

  8. Spectrographic Determination of Impurities in Ceramic Materials for Nuclear fusion Reactors. II. Analysis of Magnesium Aluminate; Determinacion Espectrografia de Impurezas en materiales Ceramicos para Fusion Nuclear. II. Analisis de Aluminato de Magnesio

    Energy Technology Data Exchange (ETDEWEB)

    Rucandio, M I; Roca, M; Melon, A

    1990-07-01

    The determination of minor and trace elements in the magnesium aluminate, considered as possible material in thermonuclear fusion reactors, has been studied. The concentration ranges are 0.1 - 0.3 % for Ca, SI and Y, and at the ppm level for Co, Cr, Fe, Hf, K, Li, Mn, Na, Ni, Se, Ta, Ti, V and Zr. Atomic emission spectroscopy with direct current are excitation and photographic detection has been employed. For Hf, Ta and Zr the use of 40% of copper fluoride as a carrier and of Nb as internal standard provide suitable sensitivities and precessions, while for the rest of elements the best results are obtained with graphite powder in different proportions and Rb or Sn as internal standard. (Author)4 refs.

  9. Development of a system based in a digital signal processor (DSP) for a simulator of power regulation in a reactor: first stage; Desarrollo de un sistema basado en un DSP para un simulador de regulacion de potencia en un reactor: 1. etapa

    Energy Technology Data Exchange (ETDEWEB)

    Benitez R, J.S.; Perez C, B. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Municipio de Ocoyoacac, 52045 Estado de Mexico (Mexico)

    2002-07-01

    The first stage of the development of a digital system based on a DSP is presented which forms part of an hybrid simulator for the power regulation in am model of the punctual kinetics of a TRIGA reactor type. The DSP performs the regulation, using a Mandami type algorithm of diffuse control. In the algorithm, the universe of the output variable is discretized for performing in an unique stage the aggregation functions and dis-diffusization. (Author)

  10. Qualification according to PDI's techniques UT EPRI methodology Phased Array for the inspection of vessels of PWR reactor with advanced robotic equipment; Cualificacion segun metodologia PDI de EPRI de te cnicas UT Phased Array para la inspeccion de vasijas de reactor PWR con eq uipos roboticos avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Gonzalez, P.; Fernandez, F.

    2014-07-01

    The techniques and procedures qualified in the program EPRI PDI are directly applicable in plants whose reference code is ASME XI - specifically the Appendix VIII-, mainly USA and countries in which it is established American PWR technology. While countries with reactors in operation technology ABB (Sweden) or type VVER (Finland and Eastern countries) requires a qualification of specific technical type ENIQ, PDI qualification is a valuable reference since it allows to deal with such qualifications with guarantees. (Author)

  11. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  14. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  15. 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.)

  16. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  17. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

    2012-07-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

  18. A method for optimization of patient dose estimation in conventional radiology; Un metodo per l'ottimizzazione della stima della dose al paziente nella radiologia tradizionale

    Energy Technology Data Exchange (ETDEWEB)

    Tofani, A.; Del Corona, A. [Azienda Unita' Sanitaria Locale 6, Livorno (Italy). Unita' Ospedaliera di Fisica Sanitaria; Niespolo, A. [Azienda Ospedaliera Pisana, Pisa (Italy). Unita' Ospedaliera di Fisica Sanitaria

    2000-05-01

    found to depend on the average body surface, a parameter which takes into account both patient height and mass. Thus, determining the normalization factor for each projection and each view allows to estimate the absorbed dose under different geometrical conditions. The method has been verified by considering four of the most common X-ray procedures (chest AP, cervical spine LAT, lumbar spine AP and head LAT). The average error on dose estimation is about 13 %. In the very next future the method will be extended to all the projections and views of ICRP Report no. 34, and we plant to integrate the described algorithm in a computer program devoted to the automatic computation of patient dose. [Italian] Il metodo raccomandato dalla International Commission on Radiological Protection (ICRP) nel suo Report n. 34(1982) per il calcolo della dose al paziente negli esami di radiodiagnostica e' basato su dati dosimetrici tabulati ottenuti mediante simulazioni Monte Carlo su fantocci antropomorfi descritti da semplici funzioni matematiche. Nel caso del calcolo della dose per un paziente adulto, le limitazioni principali di questo metodo sono due: in primo luogo i parametri geometrici dell'esame - e in particolare la distanza fuoco-pellicola e il formato della pellicola- sono fissi, e questo rende problematico l'utilizzo dei dati dosimetrici nelle condizioni effettive in cui si e' svolto l'esame, che in genere non coincideranno con quelle standard ICRP. Inoltre quando le dimensioni e la massa del paziente differiscono sensibilmente da quelle del fantoccio utilizzato nelle simulazioni (il cosidetto uomo di riferimento, di altezza pari a 174 cm e massa di 70,9 Kg) il metodo ICRP puo' portare a errori considerevoli nella stima della dose. Lo scopo del presente lavoro e' quello di indicare una possibile via di uscita per superare queste limitazioni. L'algoritmo proposto in questo lavoro si basa sull'applicazione del metodo suggerito da Huda e

  19. 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.)

  20. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  1. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  2. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  3. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

  4. A critical summary of microscopic fast-neutron interactions with reactor structural, fissile and fertile materials; Apercu critique des interactions microscopiques des neutrons rapides avec les materiaux de construction et les matieres fissiles et fertiles utilisees dans les reacteurs; Kriticheskij obzor mikroskopicheskog o vzaimodejstviya bystrykh nejtronov s konstruktsionnymi, rasshcheplyayushchimis ya i vosproizvodyashchim i reaktornymi materialami; Resumen critico de las interacciones microscopicas de los neutrones rapidos con los materiales estructurales fisionables y fertiles utilizados en los reactores

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A B [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    reactions provoquees par des neutrons rapides, en insistant sur les conditions que doivent remplir, dans les reacteurs, les donnees nucleaires fondamentales. (author) [Spanish] El autor examina el estado actual de los conocimientos acerca de las reacciones inducidas por los neutrones rapidos que se utilizan en el proyecto de reactores nucleares. Estudia con particular atencion los metodos experimentales microscopicos, sus resultados y la precision de los mismos. Considera con detalle la dispersion de los neutrones rapidos, y da los resultados de mediciones experimentales de la dispersion en el caso del oxigeno, hierro, zirconio, niobio, wolframio, torio y uranio. Expone los resultados mas significativos de los estudios experimentales de la captura de neutrones rapidos y de la fision inducida por los mismos. Las mediciones estudiadas no solo conducen a resultados de gran utilidad practica, sino que sirven como ejemplos de la aplicacion de las tecnicas nucleares experimentales mas modernas. El autor indica los terrenos en que la informacion experimental es limitada, contradictoria o inexistente. Por ultimo, formula previsiones sobre el desarrollo de los conocimientos relativos a las reacciones de los neutrones rapidos, subrayando lo referente al cumplimiento de las condiciones necesarias para que el reactor proporcione datos nucleares basicos. (author) [Russian] V doklade rassmatrivaetsya primenenie shiroko rasprostranenny kh znanij o reaktsiyakh, vyzyvaemykh bystrymi nejtronami v yadernykh proektakh reaktornykh sistem. Osnovnoe znachenie pridaetsya mikroskopicheski m ehksperimental'ny m metodam, rezul'tatam i tochnosti. Podrobno rassmatrivaetsya rasseyanie bystrykh nejtronov, vklyuchaya rezul'taty ehksperimental'nogo opredeleniya rasseyaniya na kislorode, zheleze, tsirkonii, niobii, vol'frame, torii i urane. Privodyatsya dannye, poluchennye v rezul'tate ehksperimental'ny kh issledovanij zakhvata bystrykh nejtronov i deleniya, vyzvannogo nejtronami. Izmereniya, privedennye v

  5. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  6. 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.)

  7. 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.)

  8. Manufacture of uranium compounds for research reactors fuel elements. Participation of the UCPP (Uranium compound production plant) in the Egyptian project; Elaboracion de compuestos de uranio para ser utilizados en elementos combustibles de reactores de investigacion. Participacion de la planta de fabricacion de compuestos de uranio (PFPU) en el proyecto Egipto

    Energy Technology Data Exchange (ETDEWEB)

    Boero, Norma L; Cinat, Enrique; Yorio, Daniel; Cincotta, Daniel; Ramella, Jose L; Bruno, Hernan R; Camacho, Esteban F; Pertossi, Fernando; Panunzio, Leonardo D; Fernandez, Carlos A; Sassone, Ariel [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Combustibles Nucleares

    1999-07-01

    UCPP is an international qualified supplier of U{sub 3}O{sub 8} with up to 20 % enrichment in U-235. The characteristics of this powder are those specified for fuel plates manufacture for test reactors. This paper describes the works performed in the plant since its beginning, emphasising those undertaken during the last years. The transference of U{sub 3}O{sub 8} manufacturing technology to INVAP SE, the enterprise that installed a plant of similar characteristics in the Arabian Republic of Egypt, is especially described. (author)

  9. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Yasutomi, Yoshiyuki; Nakagawa, Moroo; Sawai, Yuichi; Chiba, Akio; Suzuki, Yasutaka.

    1997-01-01

    Silicon composited with reinforcing metals is used for a divertor cooling substrate having an effect as a cooling tube to provide a silicon base composite material having increased electric resistance and toughness. The blending ratio of reinforcing materials in the form of granules, whiskers or long fibers is controlled in order to control heat conductivity, electric resistivity and mechanical performances. The divertor cooling substrate comprising the silicon base composite material is integrated with a plasma facing material. The production method therefor includes ordinary metal matrix composite forming methods such as powder metallurgy, melting penetration method, high pressure solidification casting method, centrifugal casting method and vacuum casting method. Since the cooling plate is constituted with the light metal and highly electric resistant metal base composite material, sharing force due to eddy current can be reduced, and radiation exposure can be minimized. Accordingly, a cooling structure for a thermonuclear reactor effective for the improvement of environmental problems caused by waste disposal can be attained. (N.H.)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  12. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  13. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  14. para palmito

    Directory of Open Access Journals (Sweden)

    Francisco Paulo Chaimsohn

    2007-01-01

    Full Text Available Densidades de siembra, arreglos espaciales y fertilización en pejibaye (Bactris gasipaes cv Diamantes-10 para palmito. La investigación se llevó a cabo en la Estación Experimental Los Diamantes (Guápiles, Costa Rica, el 3 de octubre del 2003, cuyo objetivo fue la evaluación del efecto de diferentes densidades de siembra (3.333, 5.000 y 6.666 plantas/ha, arreglos espaciales, y diversos métodos de fertilización (química, orgánica, sobre el crecimiento de las plantas de pejibaye para producción de palmito. Se consideraron las variables diámetro y altura del tallo primario y el número de hojas y rebrotes como indicadores de producción. El período de evolución abarcó sólo los primeros 25 meses de crecimiento en el campo. El número de hojas, la altura y el diámetro del tallo no mostraron diferencias de respuesta relevantes. Sólo el número de rebrotes disminuyó al aumentar la densidad de la población, cuando se midió a los 15 meses de edad. El efecto de la fertilización se hizo evidente después de la primera cosecha, realizada a los 20 meses, debido al aumento de la competencia entre plantas, ahora más desarrolladas. Fue entonces cuando la fertilización química indujo la producción de un mayor número y vigor de los rebrotes. Sin embargo, las prácticas evaluadas 25 meses después de la siembra, no habían infl uido hasta ese momento en el número de palmitos cosechados, ni tampoco había afectado las características físicas de los sectores foliar y caulinar del palmito.

  15. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, D.; Chavez M, C., E-mail: danmirnyi@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  16. Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants; Interfase visual para la automatizacion del pendulo instrumentado de pruebas Charpy utilizado en el programa de vigilancia de la vasija de reactores de centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A. [ININ, Carretera Mexico-Toluca Km.36.5, Mpio. de Ocoyoacac, Estado de Mexico (Mexico)]. E-mail: asrs@nuclear.inin.mx; esm@nuclear.inin.mx; jare@nuclear.inin.mx

    2004-07-01

    Inside the Programs of Surveillance of the nuclear power stations periodic information is required on the state that keep the materials with those that builds the vessel of the reactor. This information is obtained through some samples or test tubes that are introduced inside the core of the reactor and it is observed if its physical characteristics remain after having been subjected to the radiation changes and temperature. The rehearsal with the instrumented Charpy pendulum offers information on the behavior of fracture dynamics of a material. In the National Institute of Nuclear Research (ININ) it has an instrumented Charpy pendulum. The operation of this instrument is manual, having inconveniences to carry out rehearsals with radioactive material, handling of high and low temperatures, to fulfill the normative ones for the realization of the rehearsals, etc. In this work the development of a computational program is presented (virtual instrument), for the automation of the instrumented pendulum. The system has modules like: Card of data acquisition, signal processing, positioning system, tempered system, pneumatic system, compute programs like it is the visual interface for the operation of the instrumented Charpy pendulum and the acquisition of impact signals. This system shows that given the characteristics of the nuclear industry with radioactive environments, the virtual instrumentation and the automation of processes can contribute to diminish the risks to the personnel occupationally exposed. (Author)

  17. Sonochemical synthesis of a PdAg/C electrocatalyst for oxygen reduction reaction; Sintesis sonoquimica de un electrocatalizador de PdAg/C para la reaccion de reduccion de oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Godinez-Garcia, A.; Perez-Robles, J.F. [Centro de Investigacion y de Estudios Avanzados del IPN. Santiago de Queretaro, Queretaro (Mexico)]. E-mail: jperez@qro.cinvestav.mx; Solorza-Feria, O. [CINVESTAV-IPN, Mexico, D.F. (Mexico)

    2009-09-15

    The synthesis and characterization of nanocatalysts for fuel cells has been a primary line of research for the purpose of obtaining less expensive electrocatalysts with better activity. A large variety of methods exist to synthesize useful nanoparticles as electrocatalysts. Each method generates particles with a different surface morphology and, therefore, the catalytic activity usually varies depending on which is used in the synthesis. In this work, PdAg/C electrocatalysts are synthesized with high-intensity ultrasonic irradiation and compared to those obtained using a conventional method such as reduction by NaBH{sub 4}. The study of this technique is of interest because it produces highly dispersed carbon-supported nanoparticles with very clean surfaces. Each electrocatalyst was evaluated for its oxygen reduction reaction (ORR) in acid medium with cyclic voltamperometry (CV) and rotating disc electrode (RDE). The electrocatalyst was characterized with x-ray diffraction and transmission electron microscopy (TEM). The physical characterization reveals that the electrocatalyst is composed of nanometric bimetallic aggregates. An important characteristic of the PdAg/C alloy obtained using ultrasound is better activity than that obtained by reduction with NaBH{sub 4}. [Spanish] La sintesis y caracterizacion de nanocatalizadores para celdas de combustible ha sido una de las principales lineas de investigacion, con el objetivo de obtener electrocatalizadores mas baratos y con una mejor actividad. Existen una gran variedad de metodos para sintetizar nanoparticulas utiles como electrocatalizadores, cada metodo genera particulas con una morfologia superficial diferente por lo que la actividad catalitica suele variar dependiendo de cual se utilice en la sintesis. En este trabajo se sintetizan electrocatalizadores de PdAg/C con irradiacion ultrasonica de alta intensidad y se comparan con las obtenidas con un metodo convencional como es la reduccion por NaBH{sub 4}. Esta

  18. An Appraisal of Analytical Methods for Plutonium and their Applications to the Analysis of Nuclear Materials; Evaluation des Methodes Analytiques de Dosage du Plutonium et de Leur Application a l'Analyse des Matieres Nucleaires; Otsenka analiticheskikh metodov opredeleniya plutoniya i ikh primenenie dlya analiza yadernykh materialov; Metodos Analiticos de Determinacion del Plutonio y su Empleo en el Analisis de Materiales Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Milner, G. W.C.; Phillips, G. [Atomic Energy Research Establishment, Harwell, Berks. (United Kingdom)

    1966-02-15

    respectifs de la dissolution dans des melanges d'acides mineraux courants et de la dissolution faisant appel a des procedes de fusion, avec des exemples a l'appui. Les auteurs decrivent aussi les procedes utilises et les resultats obtenus, pour l'analyse des alliages Pu-U, Pu-Ce-Co et Pu-U-Mo, des oxydes et carbures de Pu-U, et des cermets de carbure de Pu-U contenant Fe, Mo et Cr. Ces matieres sont l'aboutissement de recherches metallurgiques destinees a mettre au point des combustibles nucleaires. (author) [Spanish] Para determinar el contenido de plutonio de los materiales nucleares existen diversos metodos. Si se trata de cantidades del orden del miligramo se pueden emplear la espectrofotometria diferencial por el color del Pu (III), la gravimetria basada en el PuO2{sub ,} si recuento gamma y metodos de oxidorreduccion como las valoraciones potenciometricas o amperimetricas y la culombiometria de potencial controlado. Si se trata de microgramos son preferibles el recuento alfa, la dilucion isotopica o las tecnicas polarograficas. Teniendo en cuenta que unos metodos son mas adecuados que otros para ciertos tipos de muestra, el analista tiene que resolver un dificil problema de seleccion a fin de obtener ios mejores resultados posibles. Los autores exponen las ventajas y las limitaciones puestas de manifiesto por los anos de experiencia en la A.E.R.E. y formulan observaciones acerca de la exactitud y la precision de los metodos, su sensibilidad y otras cuestiones de especial interes. Como algunos de esos metodos exigen la separacion previa del plutonio, los autores estudian el empleo de las tecnicas de intercambio anionico y de cromatografia en fase inversa, y en particular su conveniencia para el analisis de muestras radiactivas. Examinan los muchos problemas que han surgido al analizar por estos metodos, aleaciones, productos ceramicos y cermets de plutonio en diversos sistemas que contenian uranio, torio, hierro, cromo, molibdeno, cerio y cobalto. La memoria trata

  19. Neutron field characterization in the installation for BNCT study in the IEA-R1 reactor; Caracterizacao do campo de neutrons na instalacao para estudo em BNCT no reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Carneiro Junior, Valdeci

    2008-07-01

    This work aims to characterize the mixed neutron and gamma field, in the sample irradiation position, in a research installation for Boron Neutron Capture Therapy (BNCT), in the IPEN IEA-R1 reactor. The BNCT technique has been studied as a safe and selective option in the treatment of resistant cancerigenous tumors or considered non-curable by the conventional techniques, for example, the Glioblastoma Multiform - a brain cancerigenous tumor. Neutron flux measurements were carried out: thermal, resonance and fast, as well as neutron and gamma rays doses, in the sample position, using activation foils detectors and thermoluminescent dosimeters. For the determination of the neutron spectrum and intensity, a set of different threshold activation foils and gold foils covered and uncovered with cadmium irradiated in the installation was used, analyzed by a high Pure Germanium semiconductor detector, coupled to an electronic system suitable for gamma spectrometry. The results were processed with the SAND-BP code. The doses due to gamma and neutron rays were determined using thermoluminescent dosimeters TLD 400 and TLD 700 sensitive to gamma and TLD 600, sensitive to neutrons. The TLDs were selected and used for obtaining the calibration curves - dosimeter answer versus dose - from each of the TLD three types, which were necessary to calculate the doses due to neutron and gamma, in the sample position. The radiation field, in the sample irradiation position, was characterized flux for thermal neutrons of 1.39.10{sup 8} {+-} 0,12.10{sup 8} n/cm{sup 2}s the doses due to thermal neutrons are three times higher than those due to gamma radiation and confirm the reproducibility and consistency of the experimental findings obtained. Considering these results, the neutron field and gamma radiation showed to be appropriated for research in BNCT. (author)

  20. BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  1. 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

  2. Optimal restoration strategies based on heuristic techniques for electrical distribution networks; Estrategias de restablecimiento optimas basadas en tecnicas heuristicas para redes de distribucion electrica

    Energy Technology Data Exchange (ETDEWEB)

    Cruz Castrejon, J. A; Islas Perez, E; Espinosa Reza, A; Garcia Mendoza, R [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)]. E-mails: adrian.cruz@iie.org.mx; eislas@iie.org.mx; aer@iie.org.mx; rgarcia@iie.org.mx

    2013-03-15

    In this paper we present a proposed solution to the problem of finding alternatives to reset faults in radial distribution networks power systems. This solution uses a deterministic method based on the definition of heuristics and whose main objectives are to improve execution time and solution quality. This search is based on the alternate repetition of two stages: a stage that attempts to reset the unconnected areas and other areas trying ballasting overloaded. [Spanish] En este articulo se presenta una propuesta de solucion al problema de busqueda de alternativas de restablecimiento para fallas en redes de distribucion radiales en sistemas electricos de potencia. Esta solucion utiliza un metodo deterministico basado en la definicion de heuristicas y cuyos objetivos principales son: mejorar el tiempo de ejecucion y calidad de la solucion. Esta busqueda se basa en la repeticion alternada de dos etapas: una etapa que intenta restablecer las areas desconectadas y otra que intenta deslastrar las areas sobrecargadas.

  3. Special Nuclear Material Control by the Power Reactor Operator; Controle des Matieres Nucleaires Speciales par l'Exploitant d'une Centrale Nucleaire; Spetsial'nyj kontrol' nalichiya yadernykh materialov operatorom ehnergeticheskogo reaktora; Control de Materiales Nucleares Especiales por Parte de Quienes Operan el Reactor de Potencia

    Energy Technology Data Exchange (ETDEWEB)

    Cordin, R. A. [Yankee Atomic Electric Company, Boston, MA (United States)

    1966-02-15

    matieres nucleaires ne se limite pas S de simples travaux d'inventaire mais sert de base a beaucoup d'autres activites qui font partie integrante du programme d'operations de tout reacteur, par exemple les expeditions de combustible irradie, le traitement chimique du combustible epuise et la comptabilite du combustible recupere et des matieres produites au cours du fonctionnement du reacteur, et l'institution et l'application d'un regime d'assurance satisfaisant. (author) [Spanish] Combustible relativamente nuevo y sumamente valioso para la produccion de energia electrica, el uranio requiere un control muy minucioso desde el momento en que la direccion de una central asume la responsabilidad financiera inherente a su posesion hasta que como combustible parcialmente agotado se transfiere a otra instalacion en la que se recupera la parte que no se ha consumido. Antes de que se descubriera la posibilidad de emplear la energia nuclear para producir electricidad, la mayor parte de las empresas que actualmente explotan centrales nucleares explotaban centrales alimentadas con combustibles fosiles y hablan establecido sistemas de control relativamente completos y adecuados para los combustibles de ese tipo. Los responsables de las centrales nucleoelectricas deben disponer de sistemas no menos adecuados para controlar los materiales nucleares especiales que utilizan. La explotacion de los reactores de potencia no es una ciencia antigua, pero durante el tiempo relativamente corto que ha transcurrido desde que se inicio su empleo los ingenieros y hombres de ciencia han mejorado continuamente el diseflo del equipo y los metodos de trabajo con objeto de disminuir los costos de produccion y de lograr que las centrales nucleares puedan competir en el plano economico con las centrales clasicas. La administracion de los materiales nucleares debe efectuarse con metodos modernos y eficientes a fin de que los adelantos tecnologicos que han permitido reducir los costos no resulten inutiles

  4. Reconstruction of the limit cycles by the delays method; Reconstruccion de ciclos limite por el metodo de los retardos

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Calleros M, G. [CFE, CNLV, Alto Lucero, Veracruz (Mexico)]. e-mail: rcd@nuclear.inin.mx

    2003-07-01

    The boiling water reactors (BWRs) are designed for usually to operate in a stable-lineal regime. In a limit cycle the behavior of the one system is no lineal-stable. In a BWR, instabilities of nuclear- thermohydraulics nature can take the reactor to a limit cycle. The limit cycles should to be avoided since the oscillations of power can cause thermal fatigue to the fuel and/or shroud. In this work the employment of the delays method is analyzed for its application in the detection of limit cycles in a nuclear power plant. The foundations of the method and it application to power signals to different operation conditions are presented. The analyzed signals are: to steady state, nuclear-thermohydraulic instability, a non linear transitory and, finally, failure of a controller plant . Among the main results it was found that the delays method can be applied to detect limit cycles in the power monitors of the BWR reactors. It was also found that the first zero of the autocorrelation function is an appropriate approach to select the delay in the detection of limit cycles, for the analyzed cases. (Author)

  5. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  6. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  7. 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

  8. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  9. 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.)

  10. 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.)

  11. Advanced epithermal thorium reactor (AETR) physics; Physique d'un reacteur au thorium, a neutrons epithermiques, de type perfectionne (AETR); Fizika usovershenstvovannog o nadteplovogo torievogo reaktora; Fisica del reactor epitermico de tipo avanzado, alimentado con torio (AETR)

    Energy Technology Data Exchange (ETDEWEB)

    Campise, A. V. [Atomics International, Canoga Park, CA (United States)

    1962-03-15

    'etude de cet ensemble a mis en relief l'importance des donnees relatives aux sections efficaces et de l'interpretation theorique des resultats experimentaux pour l'etude d'un reacteur au thorium de type perfectionne. La precision des methodes analytiques employees a ete demontree lors de l'analyse des resultats experimentaux obtenus avec le ZPR-III. L'auteur compare trois configurations pour le transfert de chaleur, en utilisant le temps de doublement comme parametre d'optimisation. Les effets de la production de {sup 233}Pa et d'isotopes de l'uranium sur le bilan neutronique, les taux possibles de surgeneration et les caracteristiques de la combustion sont evalues en tenant compte de l'imprecision des sections efficaces nucleaires. (author) [Spanish] El autor estudia la concepcion del reactor AETR desde el punto de vista de la teoria actual de los parametros nucleares y del balance neutronico. En los sistemas moderados por grafito examina el efecto de la captura por resonancia en el torio para energias medias de absorcion del orden de 0,10 a 100 keV. Aplica formulas de resonancia angosta y de resonancia ancha para obtener la integral de resonancia efectiva en funcion de la temperatura, correspondiente a las barras de torio, y dicho parametro se expresa como secciones eficaces equivalentes de varios grupos. Se ha disenado y construido un conjunto critico para obtener datos nucleares indispensables en la gama de energias intermedias. En el diseno nuclear de dicho conjunto, se ha tenido particularmente en cuenta la importancia de los datos relativos a secciones eficaces y la interpretacion teorica de estos resultados experimentales, cosas ambas relacionadas con el diseno del reactor AETR. La precision de los metodos analiticos ha quedado demostrada por el estudio de los resultados experimentales obtenidos con el reactor ZPR-III. Se comparan tres sistemas de transmision de calor utilizando el tiempo de duplicacion como parametro optimo. Se estudia el efecto de la formacion

  12. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  13. 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

  14. 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.

  15. Data Evaluation Problems in the Pulsed Neutron Source Method; Problemes d'Evaluation des Donnees dans les Applications de la Methode de la Source Pulsee; Problemy otsenki dannykh pri primenenii metoda istochnika impul'snykh nejtronov; Problemas de Evaluacion de Datos en el Metodo de la Fuente de Neutrones Pulsados

    Energy Technology Data Exchange (ETDEWEB)

    Pal, L.; Bod, L.; Szatmary, Z. [Central Research Institute for Physics, Hungarian Academy of Sciences, Budapest (Hungary)

    1965-08-15

    moyen de transport dans de nombreuses matieres hydrogenees. (author) [Spanish] Desde hace mas de diez anos se viene empleando el metodo de los neutrones pulsados para determinar los parametros de difusion de los neutrones termicos. Este metodo ha resultado especialmente idoneo para el estudio de los moderadores hidrogenados. Los autores dan los resultados de las mediciones hechas en agua, benceno, tolueno, xileno, ciclohexano, n-hexano y difenilo, y examinan los metodos para evaluar con el minimo error los parametros de difusion de interes partiendo de los datos obtenidos en las mediciones. Se examina detenidamente, tanto en el plano teorico como en el experimental, el efecto de tiempo de vuelo que se manifiesta por razones tecnicas cuando se investigan moderadores a elevadas temperaturas, y que influye en el valor de la constante de desintegracion, y se exponen las condiciones geometricas en que se ha de operar para obtener resultados satisfactorios. Los autores determinan tambien el efecto que sobre la constante de desintegracion ejerce el tiempo muerto del detector neutronico y del amplificador, y demuestran que es despreciable cuando [i(t) {tau}]{sup 2} << 1, siendo i/t/dt el numero de neutrones detectados en el intervalo de tiempo/t, t + dt/, y {tau} el tiempo muerto. Se evaluo la constante de desintegracion empleando el metodo de la maxima probabilidad. Para determinar inequivocamente la funcion de probabilidad, se eligio un analizador multicanal que acumulaba como maximo una sola senal por ciclo de analisis y por canal. Se comprobo que la incorporacion del parametro caracteristico del fondo a la funcion de probabilidad tiene primordial importancia. Los parametros de la funcion de maxima probabilidad se evaluaron en una calculadora aplicando el metodo de iteracion de Newton. Cuando no se tiene debidamente en cuenta el fondo, este puede causar una mayor contaminacion aparente de armonicos, la cual, a su vez, puede interpretarse como un fondo aparente. Por ello es

  16. Radioactive Waste Facilities at the Australian Atomic Energy Commission Research Establishment; Installations pour le Traitement des Dechets Radioactifs au Centre de Recherche de la Commission Australienne a l'Energie Atomique; 0423 0421 0422 0414 ; Dispositivos para Evacuacion de Desechos Radiactivos en el Centro de Investigaciones de la Australian Atomic Energy Commission

    Energy Technology Data Exchange (ETDEWEB)

    Berglin, C. L.W.; Keher, L. H.; Miles, G. L.; Wilson, A. R.W. [Australian Atomic Energy Commission Research Establishment, Lucas Heights, Sydney, NSW (Australia)

    1960-07-01

    'impuretes fertiles (appliquant un procede calcium-fer-phosphate) et de bassins de retenue. Le memoire examine les diverses methodes de concentration des impuretes et de traitement secondaire qui sont actuellement a l'etude. Les auteurs etudient la formule d'evacuation et la dilution probable dans le Woronora, et decrivent un essai de dilution qui a ete effectue dans les eaux touchees par la maree. On se propose d'enfouir, le cas echeant apres empaquetage, tous les dechets solidesi de faible activite. Les auteurs etudient le choix et l'emplacement de la zone d'elimination. Ils font un expose sur la construction d'une installation destinee au stockage et a l'elimination de dechets solides a activite elevee. On envisage l'e vaporation et le stockage des dechets liquides a activite moyenne et elevee. Le memoire donne des renseignements detailles sur les depenses d'equipement et d'exploitation afferentes a l'usine de traitement des effluents et a d'autres installations pour le traitement des dechets. (author) [Spanish] La memoria describe las instalaciones previstas para la recogida, tratamiento y evacuacion de los desechos radiactivos en Lucas Heights, en relacion con las cantidades que se produciran segun los calculos. Los efluentes de baja actividad se dividen en tres tipos: a) Aguas de las cloacas, b) Desechos de funcionamiento, provenientes de la torre de refrigeracion del reactor, de los talleres mecanicos y otras zonas no activas, c) Efluentes de los laboratorios y otras zonas activas. La planta de tratamiento para los efluentes del ultimo tipo consiste en principio en tanques de mezcla y dosificacion de alcali, un clarificador de lodos (que emplea un sistema a base de calcio - herro - fosfato) y tanques receptores. Se estan estudiando metodos para concentrar los lodos y metodos de tratamiento secundario, que se discuten en la memoria. La memoria estudia la forma de descarga y el grado de dilucion que se espera obtener en el rio Woronora, asi como un experimento de dilucion que se

  17. Influência do tempo de detenção hidráulica em um sistema UASB seguido de um reator biológico com fungos para tratar efluentes de indústria de castanha de caju Influence of the time of detention hidraulic of a sistem UASB followed by a biological reactor with fungi to treat efluent of cashew nut industry

    Directory of Open Access Journals (Sweden)

    Emília Maria Alves Santos

    2006-03-01

    Full Text Available Nesta pesquisa, estudou-se a influência do tempo de detenção hidráulica (TDH em um sistema constituído de um reator anaeróbio tipo UASB seguido de um reator biológico com fungos (RBF para tratar efluente de uma indústria de beneficiamento de castanha de caju. O presente trabalho foi dividido em uma fase de fluxo descontínuo (batelada e uma fase de fluxo contínuo (UASB - RBF, que constituiu-se de sete etapas ( 8h e 2h, 8h e 1h, 4h e 8h, 4h e 6h, 4h e 4h, 4h e 2h e 4h e 1h, onde foi avaliada a influência do TDH na remoção de: DQO (Demanda Química de Oxigênio, amônia, nitrato e ortofosfato. Uma combinação que apresentou melhores resultados, foi a etapa de 4h (TDH do reator UASB e 2h (TDH do RBF, apresentando remoções de: 93,8% de DQO, 86,7% de nitrato, 38,3% de amônia e 16% de ortofosfato.In this research, it was studied the effect of hydraulic retention time (HRT in a system comprised of an Upflow Anaerobic Sludge Blanket (UASB reactor and a Biological Reactor with Fungi (BRF for treatment of the efluent of the industry of cashew nut improvement. The work was divided in two phases: batch reactors using shaking flasks and continuous-feed reactors (UASB-BRF. The UASB reactor was operated at HRT of 4 and 8 h, whereas the BRF was operated at HRT varying from 1 to 8 h. The performance of both reactors was evaluated based on the removal efficiency of chemical oxygen demand (COD, ammonium, nitrate, and orthophosfate. The results show that the best results were achieved when the UASB was operated at HRT of 4 h and the BRF was operated at HRT of 2 h, when the system removed 93,8% of the COD, 86,7% of the nitrate, 38,3% of the ammonium and 16% of the orthophosfate.

  18. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  19. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  20. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  1. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  2. Control for nuclear reactor

    International Nuclear Information System (INIS)

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  3. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  4. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  5. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  6. continuous stirred tank reactor (CSTR)

    African Journals Online (AJOL)

    AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... stirred tank reactor (CSTR) and the small and large intestines as plug flow reactor (PFR) ... from the two equations are used for the reactor sizing of the modeled reactors.

  7. Application of the identification methods from Hilbert and Prony to the study of oscillatory phenomena in electrical power systems; Aplicacion de los metodos de identificacion de Hilbert y Prony al estudio de fenomenos oscilatorios en sistemas electricos de potencia

    Energy Technology Data Exchange (ETDEWEB)

    Andrade Soto, Manuel Antonio

    2002-10-15

    In the present thesis work the application of identification techniques is investigated based on methods of spectral analysis to the study of the instantaneous characteristics of signals obtained by means of digital simulation of the dynamic behavior of the power system. The study focuses on the perspectives developed from two different approaches of analysis: the use of lineal methods of spectral analysis and the use of methods of non-lineal analysis, based on the concept of an analytical signal. The developed tools are applied to the study of two phenomena of electromechanical origin of different characteristics in complex power systems. A comparison between the results obtained is performed for these techniques and the possibility of its application is discussed for the problem of on-line identification in power systems. [Spanish] En el presente trabajo de tesis se investiga la aplicacion de tecnicas de identificacion basadas en metodos de analisis espectral al estudio de las caracteristicas instantaneas de senales obtenidas mediante simulacion digital del comportamiento dinamico del sistema de potencia. El estudio se centra en las aproximaciones desarrolladas desde dos enfoques distintos de analisis: la utilizacion de metodos lineales de analisis espectral y la utilizacion de metodos de analisis no lineal, basados en el concepto de una senal analitica. Las herramientas desarrolladas se aplican al estudio de dos fenomenos de origen electromecanico de caracteristicas distintas en sistemas complejos de potencia. Se hace una comparacion entre los resultados obtenidos por estas tecnicas y se discute la posibilidad de su aplicacion al problema de identificacion en linea en sistemas de potencia.

  8. 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)

  9. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  10. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  11. Reactor utilization, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1984-01-01

    Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

  12. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  13. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  14. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

  15. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  16. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  17. Rotating reactors : a review

    NARCIS (Netherlands)

    Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

    2013-01-01

    This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

  18. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  19. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is,