Development of technologies for nuclear reactors of small and medium sized

International Nuclear Information System (INIS)

2011-08-01

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.

Estudio hidrodinámico de reactores empacados de flujo ascendente(REFA)

OpenAIRE

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

2014-01-01

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

Operating Experience with the VERA Zero-Energy Fast Reactor; Fonctionnement du Reacteur VERA a Neutrons Rapides, de Puissance Zero; Opyt ehkspluatatsii reaktora VERA na bystrykh nejtronakh nulevoj moshchnosti; Experiencia Adquirida con el Reactor Rapido VERA de Potencia Nula

Energy Technology Data Exchange (ETDEWEB)

Weale, J. W.; McTaggart, M. H.; Goodfellow, H.; Paterson, W. J. [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

1964-02-15

elements combustibles; incidences des irregularites des faces et de l'heterogeneite de la structure sur l'exactitude avec laquelle on determine la masse critique d'un assemblage equivalent 'ideal'; possibilite de reproduire l'etat critique apres avoir demonte l'assemblage ou separe les deux moities; variation de la reactivite lors de la separation des deux moities, et effets d'un chargement asymetrique; sensibilite de divers compteurs, intensite de la source de neutrons, utilisation d'un accelerateur comme source de neutrons; vitesse de reponse des circuits de securite et restrictions qu'elle peut entrainer pour la vitesse d'assemblage des deux moities; precautions supplementaires qu'il faut prendre pour utiliser le plutonium comme combustible; notes sur l'exactitude des mesures de la reactivite et sur les limitations d'ordre pratique imposees a diverses autres mesures en physique des reacteurs. (author) [Spanish] Los autores describen brevemente las caracteristicas de un reactor rapido de potencia nula de dos mitades, en especial las que determinan la posibilidad de efectuar mediciones de precisidn en la esfera de la fisica de los reactores. Examinan las ventajas e inconvenientes de este tipo de reactor, basandose en la experiencia adquirida durante dos anos de funcionamiento. Exam in an los siguientes temas: ventajas, desde el punto de vista experimental, de la disposicion y de la construccidn en dos mitades; dimensiones y tolerancias de los elementos combustibles, y precisian con la que se puede determinar la posicion de dichos elementos; influencia de las irregularidades marginales y de heterogeneidad estructural sobre la exactitud con la que se determine la masa critica de un conjunto 'ideal' equivalente; posibilidad de reproducir las condiciones criticas despues de desmontar el conjunto o de separar las dos mitades; variaciones de la reactividad cuando se separan las dos mitades, en especial los efectos de una carga asimetrica; sensibilidad de varios contadores

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

OpenAIRE

Zambrano Flores, Johanna Vanessa

2013-01-01

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

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

Directory of Open Access Journals (Sweden)

Luis Eduardo Forero

2004-01-01

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

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

OpenAIRE

Cuevas-Rodríguez, Germán; Tejero Monzón, Iñaki

2003-01-01

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

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.

Operating Experience in Nuclear Power Plants with Boiling-Water Reactors; Experience acquise dans l'exploitation des reacteurs a eau bouillante; Opyt ehkspluatatsii kipyashchago reaktora; Experiencia adquirida con la explotacion de reactores de agua hirviente

Energy Technology Data Exchange (ETDEWEB)

Ascherl, R. J. [General Electric Company, San Jose, CA (United States)

1963-10-15

dans la centrale et aux alentours restent nettement au-dessous des limites fixees par les permis d'exploitation. La simplicite et la facilite d'exploitation des reacteurs a eau bouillante se sont confirmees. La rapidite de reponse du reacteur de Dresden aux variations de la demande est excellente. Aucun des travaux d'entretien et de reparation, quelle que soit leur importance, n'exige de qualification speciale ni ne comporte de risques de radioexposition excessifs. L'inspection generale et la revision de la turbine de la centrale de Dresden n'ont revele aucun probleme d'entretien apres 12 000 h de fonctionnement avec la vapeur en cycle direct. Or, on avait continue l'exploitation apres avoir constate que des cartouches de combustible etaient deteriorees. (author) [Spanish] El autor sefiala que la experiencia adquirida con la explotacion de centrales nucleoeleonicas que utilizan reactores de agua hirviente es ya bastante considerable. En efecto, al finalizar el aflo 1962 se habian generado mas de 2,2.10{sup 9} kWh en tres centrales nucleoelectricas integradas en redes de distribucion: la central nucleoelectrica de Dresden (Commonwealth Edison Company, Moriis, Illinois), la central nucleoelectrica de Vallecitos (Pacific Gas and Electric Company, Pleasanton, California) y la central nucleoelectrica de Kahl (Reinisch-Westfalisches Elektrizitatswerk y Bayernwerk A.G., Kahl-am-Main, Republica Federal de Alemania). El rendimiento de estas centrales explotadas en condiciones normales de produccion de electricidad es excelente. Los factores de disponibilidad y de capacidad de los reactores y de las plantas constituyen una base firme para prever que las centrales nucleoelectricas dotadas de reactores de agua hirviente funcionaran a regimen continuo en condiciones de seguridad. En 1963 entraran en servicio cuatro nuevas centrales equipadas con reactores de agua hirviente: la central nucleoelectrica de Big Rock Point (Consumere Power Company, Charlevoix, Michigan), la central

Riego subsuperficial con aguas residuales tratadas

OpenAIRE

Crespi, Raúl; Camacho, Emilio; Polo, María José

2009-01-01

La reutilización de aguas residuales tratadas para riego por goteo subsuperficial (RGS) de cultivos extensivos, es una práctica reconocida mundialmente, siendo la obturación de emisores un serio problema en la operación de estos sistemas. Durante el ciclo agrícola 2002/03, se realizó un ensayo experimental usando dos tipos de agua: limpia (AL) y residual tratada (AR) en un sistema mixto combinando un reactor biológico con laguna de maduración y tres tipos de protectores del emisor: micro dosi...

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-12-31

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

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

Directory of Open Access Journals (Sweden)

Rangel Jara Hermes Augusto

1997-01-01

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

Productivity of a nuclear chemical reactor with gamma radioisotopic sources; Rendimiento de un reactor quimico-nuclear con fuentes radioisotopicas gamma

Energy Technology Data Exchange (ETDEWEB)

Anguis T, C

1975-07-01

According to an established mathematical model of successive Compton interaction processes the made calculations for major distances are extended checking the acceptability of the spheric geometry model for the experimental data for radioisotopic sources of Co-60 and Cs-137. Parameters such as the increasing factor and the absorbed dose served as comparative base. calculations for the case of a punctual source succession inside a determined volume cylinder are made to obtain the total dose, the deposited energy by each photons energetic group and the total absorbed energy inside the reactor. Varying adequately the height/radius relation for different cylinders, the distinct energy depositions are compared in each one of them once a time standardized toward a standard value of energy emitted by the reactor volume. A relation between the quantity of deposited energy in each point of the reactor and the conversion values of chemical species is established. They are induced by electromagnetic radiation and that are reported as ''G'' in the scientific literature (number of molecules formed or disappeared by each 100 e.v. of energy). Once obtained the molecular performance inside the reactor for each type of geometry, it is optimized the height/radius relation according to the maximum production of molecules by unity of time. It is completed a bibliographical review of ''G'' values reported by different types of aqueous solutions with the purpose to determine the maximum performance of molecular hydrogen as a function of pH of the solution and of the used type of solute among other factors. Calculations for the ethyl bromide production as an example of one of the industrial processes which actually work using the gamma radiation as reactions inductor are realized. (Author)

Productivity of a nuclear chemical reactor with gamma radioisotopic sources; Rendimiento de un reactor quimico-nuclear con fuentes radioisotopicas gamma

Energy Technology Data Exchange (ETDEWEB)

Anguis T, C

1975-07-01

According to an established mathematical model of successive Compton interaction processes the made calculations for major distances are extended checking the acceptability of the spheric geometry model for the experimental data for radioisotopic sources of Co-60 and Cs-137. Parameters such as the increasing factor and the absorbed dose served as comparative base. calculations for the case of a punctual source succession inside a determined volume cylinder are made to obtain the total dose, the deposited energy by each photons energetic group and the total absorbed energy inside the reactor. Varying adequately the height/radius relation for different cylinders, the distinct energy depositions are compared in each one of them once a time standardized toward a standard value of energy emitted by the reactor volume. A relation between the quantity of deposited energy in each point of the reactor and the conversion values of chemical species is established. They are induced by electromagnetic radiation and that are reported as ''G'' in the scientific literature (number of molecules formed or disappeared by each 100 e.v. of energy). Once obtained the molecular performance inside the reactor for each type of geometry, it is optimized the height/radius relation according to the maximum production of molecules by unity of time. It is completed a bibliographical review of ''G'' values reported by different types of aqueous solutions with the purpose to determine the maximum performance of molecular hydrogen as a function of pH of the solution and of the used type of solute among other factors. Calculations for the ethyl bromide production as an example of one of the industrial processes which actually work using the gamma radiation as reactions inductor are realized. (Author)

Analysis of Organizational Structure in Road Transport Organizations from Spanish Atlantic-Mediterranean Region; Analisis de la Estructura Organizativa de las Empresas del Sector del Transporte por Carretera en el Arco Atlantico-Mediterraneo Espanol

Energy Technology Data Exchange (ETDEWEB)

Gamero, N; Sola, R

2009-11-25

This report illustrates the main results of a research on organizational structure in road transport organizations from Spanish Atlantic-Mediterranean region. This study is included in the project Potenciacion de la competitividad del tejido empresarial espanol a traves de la logistica como factor estrategico en un entorno global (GLOBALOG) supported by the Spanish Ministry of Education and Science in 2007. The main aim was to increase the level of competitiveness of this type of companies developing new knowledge, methodologies and practices. In order to analyse organizational structure, we used Minstbergs (2003) typology. The results will provide indicators to improve organizational objectives. (Author) 14 refs.

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

International Nuclear Information System (INIS)

Hien, P.D.

1999-01-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

Energy Technology Data Exchange (ETDEWEB)

Hien, P.D. [Vietnam Atomic Energy Agency, Hanoi (Viet Nam)

1999-08-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

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...

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...

Análisis experimental y computacional de modelos termofluidomecánicos para el diseño de ingeniería con metales líquidos

OpenAIRE

Peña Bandrés, Alberto

2013-01-01

Es importante disponer de una herramienta con la cual diseñar dispositivos de uso industrial y comercial que trabajen con metales líquidos (fuentes de neutrones de alta intensidad, núcleos de sistemas de transmutación nuclear, reactores de fisión de nueva generación, instalaciones de irradiación de materiales o reactores de fusión nuclear). Los códigos CFD (Computational Fluid Dynamics) son una de esas herramientas, y la manera de llevar a cabo su validación es la simulación de experimentos e...

An economic evaluation of the economics of the Rancho Seco nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Craig, P P [California Univ., Davis, CA (USA). Dept. of Applied Science; Marcus, W B [JBS Energy, Inc., Broderick, CA (USA)

1991-03-01

Escalating costs of operating reactors may make it economically advantageous to close them down and purchase small, cheap technologies such as combined-cycle systems. We examine the arguments pro and con for the Rancho Seco nuclear reactor, owned by the Sacramento Municipal Utility District (SMUD) and now permanently shut-down. We conclude that if the reactor could be run no better than it has run in the past, there was no clear advantage to continued operation. Optimistic scenarios show a net advantage to running it, and pessimistic scenarios show a net disadvantage. The total range of plausible costs is narrow; the financial impact of either choice on consumers would have been small. The analysis suggests that decisions on whether to close down existing reactors are highly case specific, but that strong economic arguments for shut-down can exist if reactor performance is poor and/or if maintenance expenses are high. (author).

An economic evaluation of the economics of the Rancho Seco nuclear reactor

International Nuclear Information System (INIS)

Craig, P.P.

1991-01-01

Escalating costs of operating reactors may make it economically advantageous to close them down and purchase small, cheap technologies such as combined-cycle systems. We examine the arguments pro and con for the Rancho Seco nuclear reactor, owned by the Sacramento Municipal Utility District (SMUD) and now permanently shut-down. We conclude that if the reactor could be run no better than it has run in the past, there was no clear advantage to continued operation. Optimistic scenarios show a net advantage to running it, and pessimistic scenarios show a net disadvantage. The total range of plausible costs is narrow; the financial impact of either choice on consumers would have been small. The analysis suggests that decisions on whether to close down existing reactors are highly case specific, but that strong economic arguments for shut-down can exist if reactor performance is poor and/or if maintenance expenses are high. (author)

preescolares desnutridos con madres con obesidad y sin obesidad

Directory of Open Access Journals (Sweden)

Viridiana Vanessa Conzuelo-González

2009-01-01

Full Text Available El primer objetivo fue conocer cuántos menores de cinco años con diferentes grados de desnutrición tienen una madre con sobrepeso/obesidad/ en una comunidad indígena que vive en extrema pobreza y bajo condiciones de migración masculina internacional. El segundo fue comparar tres variables socionutricionales (ingreso familiar, educación de la madre y adecuación nutrimental de la dieta diaria entre estos hogares y los hogares con desnutrición infantil y madres sin obesidad. Se realizó un estudio transversal (2006-2007, en la comunidad mazahua de San Francisco Tepeolulco, Municipio de Temascalcingo; que incluyó a 85 hogares integrados por preescolares con desnutrición inscritos al programa Oportunidades. Se determinó el estado nutrición de los preescolares con indicadores antropométricos y se obtuvo el IMC de las madres de estos infantes. Se aplicó una encuesta socionutricional, incluida el recordatorio de 24 horas, y complementado con la observación participante (cualitativa. Se encontró que 83% de las madres mazahuas presentaron sobrepeso u obesidad. El estado de nutrición de los preescolares con madres con obesidad presentó un porcentaje mayor de desnutrición (76%. En la variable género, se encontró que 54% de los niños con madres con obesidad tenía baja talla. Al relacionar el nivel educativo de la madre, esta variable resultó ser estadísticamente significativa (p=0.015, donde el analfabetismo está más relacionado con la desnutrición infantil que tienen madres de bajo y/o peso normal. La elevada prevalencia de hogares conformados con preescolares con desnutrición y madres con obesidad, es un síntoma más de la pobreza en zonas indígenas en México, con bajo índice de desarrollo humano.

How much does a tokamak reactor cost?

Science.gov (United States)

Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

2017-10-01

The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

Compatibility of sodium with ceramic oxides employed in nuclear reactors; Compatibilidad del sodio con oxidos ceramicos utilizados en reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Acena Moreno, V

1981-07-01

This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

Science.gov (United States)

Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

2018-08-01

Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.

Design of the core of a breed/burn fast reactor with the deterministic code KANEXT; Diseno del nucleo de un reactor rapido de cria/quemado con el codigo deterministico KANEXT

Energy Technology Data Exchange (ETDEWEB)

Lopez S, R. C.; Francois L, J. L., E-mail: rcarlos.lope@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

2014-10-15

The breeding fast reactors are interesting because they generate more plutonium than they consume, however, the fuel has to be reprocessed for the generated plutonium is used in another reactor. In a breed/burn reactor (BBR) the plutonium is generated and used -in situ- inside the same reactor, reducing this way costs and the proliferation possibility. In this work, the core of a BBR was designed; cooled by sodium that consists of 210 active assemblies and 7 spaces for control rods, each assembly consists of 169 pines. The design differs from other BBR it includes a blanket in the reactor center. The above-mentioned was to take advantage of the fact by geometry that the population of fast and epithermal neutrons will be high in the area, due to the fissions in adjacent fissile areas. Favorable results were obtained, although not definitive with exchange scheme of spent fuel. Efforts should be made in the future to homogenize the power generation within the reactor and replace the spent assemblies more efficiently. (Author)

Correction method for critical extrapolation of control-rods-rising during physical start-up of reactor

International Nuclear Information System (INIS)

Zhang Fan; Chen Wenzhen; Yu Lei

2008-01-01

During physical start-up of nuclear reactor, the curve got by lifting the con- trol rods to extrapolate to the critical state is often in protruding shape, by which the supercritical phenomena is led. In the paper, the reason why the curve was in protruding was analyzed. A correction method was introduced, and the calculations were carried out by the practical data used in a nuclear power plant. The results show that the correction method reverses the protruding shape of the extrapolating curve, and the risk of reactor supercritical phenomena can be reduced using the extrapolated curve got by the correction method during physical start-up of the reactor. (authors)

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

Study of the neutronic behavior of a fuel assembly with gadolinium of a reactor HPLWR; Estudio del comportamiento neutronico de un ensamble combustible con gadolinia de un reactor HPLWR

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. (Mexico)

2012-10-15

This work presents a neutronic study of a square assembly design of double line of fuel rods, with moderator box to center of the arrangement, for the nuclear reactor cooled with supercritical water, High Performance Light Water Reactor (HPLWR). For the fuel analyses of the reactor HPLWR the neutronic code Helios-2 was used, settling down as the first study on fuel under conditions of supercritical water that has been simulated with this code. The analyzed variables, essentials in the neutronic design of any reactor, were the infinite neutrons multiplication factor (k{infinity}) and the maximum power peaking factor (PPF{sub max}), as well as the reactivity coefficients by the fuel temperature. The k{infinity} and PPF{sub max} values were obtained under conditions in cold (293.6 K) and in hot (to 880.8 K). The tests were realized for a reference fuel assembly design, with 40 fuel rods with enrichments of 4 and 5% of U-235, and considering different concentrations of consumable poison (gadolinium - Gd{sub 2O3}) in some rods of the same assembly. The obtained results show values k{infinity} and PPF{sub max} minors to the present in the conventional light water reactors. Moreover, the reactivity coefficients by fuel temperature were verified with the purpose of satisfying the safety conditions required in the nuclear reactors. (Author)

Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

Energy Technology Data Exchange (ETDEWEB)

Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

Fuel burn analysis of a sodium fast reactor with KANEXT and Serpent; Analisis de quemado de combustible de un reactor rapido de sodio con KANEXT y SERPENT

Energy Technology Data Exchange (ETDEWEB)

Lopez S, R. C.; Francois L, J. L., E-mail: rcarlos.lope@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

2015-09-15

The fast reactors cooled by sodium are one of the options considered in the Generation IV. Since most of the reactors of Fourth Generation are still in development stage, is necessary to have efficient and reliable computational tools, this in order to obtain accurate results in reasonable computational times. In this paper is introduced and describes the deterministic code KANEXT (KArlsruhe Neutronic EXtended Tool) and is compared against a Monte Carlo code of more diffusion: Serpent. KANEXT, being a modular code requires the interaction of different modules to perform a job, this interaction of modules is described in this article. The parameters to be compared are the results of the neutron multiplication effective factor and the evolution of isotopes during the burning. The mentioned comparison is carried out for a fast reactor cooled by sodium of relatively small size compared to commercial size reactors. In this paper the particularities of the reactor are described, important for the analysis such as geometry, enrichments, reflector, etc. The considerations in the implementation in both codes are also described, as are simplifications, length of the burning steps, possible solutions of the Bateman equations for the burning fuel in Serpent and the solution options for transport (P3) and diffusion (P1) in KANEXT. The results show good correspondence between Serpent and KANEXT, which give confidence to continue using KANEXT as the main tool. Respect to computation time, time saving is evident with the use of deterministic codes instead of Monte Carlo codes, in this particular case, the time savings using KANEXT is about 98.5% of the time used by Serpent. (Author)

Optimization of SFR Reactor design with recycling or minor actinides; Optimizacion del diseno de reactor SFR con reciclado de actinidos minoritarios

Energy Technology Data Exchange (ETDEWEB)

Martin-Fuertes, F.; Vazquez, M.; Alvarez, F.

2012-07-01

In this paper we show results of the design features and ESFR optimized in three configurations: the reference, load the minority actinides homogeneous throughout the reactor and the high content of AM on a radial mantle. Was calculated reactivity evolution in five cycles burned (2050 days) to recharge One approach. To do this, we have employed EVOLCODE2 a development tool of CIEMAT own coupling MCNPX and ORIGEN.

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

OpenAIRE

Mendoza Sépulveda, César Camilo

2013-01-01

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

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

Influencia de la microestructura en el comportamiento a fatiga de aceros inoxidables austeníticos con alto contenido en molibdeno

Directory of Open Access Journals (Sweden)

Oñoro, J.

2006-02-01

Full Text Available Austenitic stainless steels with molybdenum present high mechanical properties and corrosion resistance to aggressive environments. These steels have been used to tank and vessel components for high corrosive liquids as phosphoric, nitric and sulphuric acids. These materials with low carbon and nitrogen addition have been proposed candidates as structural materials for the international thermonuclear experimental reactor (ITER in-vessel components. Molybdenum addition in austenitic stainless steel improves mechanical and corrosion properties, but with it can produce the presence of nitrogen microstructure modifications by presence or precipitation of second phases. This paper summarises the fatigue and corrosion fatigue behaviour of two 317LN stainless steels with different microstructure. Fully austenitic steel microstructure show better fatigue, corrosion fatigue resistance and better ductility than austenitic steel with delta ferrite microstructure, mainly at low stresses.

Los aceros inoxidables austeníticos con elevados contenidos en molibdeno presentan alta resistencia mecánica y resistencia a los medios corrosivos. Se utilizan en la construcción de depósitos y recipientes para el almacenamiento y transporte de líquidos altamente corrosivos, tales como ácido fosfórico, nítrico o sulfúrico. Estos materiales con bajo carbono y adiciones de nitrógeno han sido propuestos como candidatos para materiales estructurales en la fabricación de la vasija del reactor experimental termonuclear internacional (ITER. La adición de molibdeno mejora las propiedades frente a la corrosión de los aceros inoxidables austeníticos. Sin embargo, este aumento del contenido en molibdeno, junto con la presencia de nitrógeno, puede producir modificaciones microestructurales, por la aparición de fases precipitadas o segundas fases. En este trabajo, se analiza el comportamiento la fatiga y corrosión-fatiga de dos aceros inoxidables austen

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

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.)

con dietas suplementadas con Cromo-L-metionina

Directory of Open Access Journals (Sweden)

Ram\\u00F3n Garc\\u00EDa-Castillo

2006-01-01

Full Text Available Un total de 48 cerdos (Sus scrofa domesticus; 24 machos castrados y 24 hembras cruzados (Yorkshire, Hampshire, Duroc y Landrace de 3,5 a 4,0 meses de edad y 60,0 ± 5,0 kg PV en finalización. Se alimentaron con dietas isoproteícas (14,5 % PC e isoenergéticas (3.400 kcal EM/kg de MS, adicionadas con Cr-L-metionina (MiCroPlex® (0, 200, 400 y 600 ppb. El experimento tuvo una duración de 45 días y se realizó de agosto a noviembre del 2002 en las instalaciones de la Universidad Autónoma Agraria Antonio Narro, localizada en Saltillo, Coahuila, México. Al tener los animales aproximadamente 95 kg PV, se tomó muestra de 15 ml de sangre por cada animal para determinar la concentración de glucosa, ácido úrico, creatinina, urea, proteinas totales y colesterol. Se aplicó un diseño completamente al azar con arreglo factorial 2 x 4; dos para el factor sexo y cuatro para nivel de cromo. Los metabolitos en suero no fueron afectados (P>0,05 por el factor sexo. La glucosa en suero disminuyó (P<0,05 y el colesterol incrementó (P<0,05 con cromo en la dieta. Se concluye que el Cr incrementa el metabolismo de glucosa y disminuye el de colesterol, con lo cual puede haber energía disponible para síntesis de proteína la cual es necesaria para el crecimiento de los animales

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)

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.)

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

A review of investigations on wastewater treatment with MSOBR (membrane supported and oxygenated biofilm reactors); Una revision de las investigaciones sobre el tratamiento de aguas residuales con RBSOM (reactores de biopelicula que emplean membranas con material soporte y medio de oxigenacion)

Energy Technology Data Exchange (ETDEWEB)

Esteban Garcia, A. L.; Tejero Monzon, I.

2007-07-01

MSOBR (membrane supported and oxygenated biofilm reactors) are biological reactors for wastewater treatment in which biofilm support and oxygenation functions are carried out by gas permeable membranes. In these conditions, with oxygen and substratum (carbonaceous, nitroge neous) diffusing into the biofilm from opposite sides, different environments are developed inside the biofilm, allowing simultaneous nitrification, denitrification and carbon removal. Other added advantages, such us the possibility of a high oxygen transfer efficiency or those derived from the absence of bubbles in aeration (minimizing foaming and VOC emissions), have lead numerous research groups to work in the development of different MSOBR systems, with promising results that make possible to consider their practical applicability in the near future. (Author) 69 refs.

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

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.)

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

Activación del topacio natural irradiado por neutrones en el núcleo del reactor RP-10

OpenAIRE

Gómez, J.; Parreño, Fernando; Lázaro, Gerardo; Vela, Mariano

2003-01-01

Se obtuvieron cristales de topacio activados al ser irradiados con neutrones dentro del núcleo del reactor RP-10. La activación depende del flujo de neutrones, por ello se desarrolló portamuestras (canes de irradiación) para absorber que son los causantes de la activación

Tratamiento con implantes Leader-Nano en paciente con oligodoncia

Directory of Open Access Journals (Sweden)

Salvador Javier Santos Medina

2015-03-01

Full Text Available Los implantes dentales de titanio han revolucionado el mundo de la rehabilitación desde su surgimiento. De manera particular, el empleo de implantes de carga inmediata acorta el tiempo quirúrgico y protésico, con el consiguiente bienestar estético. Se presenta el caso de una paciente femenina de 32 años de edad, con antecedentes de oligodoncia de ambos incisivos laterales superiores y portadora de prótesis parcial acrílica. Fue atendida por el equipo multidisciplinario de implantes en la Clínica Estomatológica Docente “3 de Octubre” y se le realizó tratamiento de rehabilitación integral con implantes Leader-Nano y prótesis fija con corona acrílica sobre dichos implantes. La implantología fue satisfactoria en la paciente; la mejoría estética y funcional, así como la satisfacción de la paciente, fueron los principales logros obtenidos

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Experiencias de haber crecido con un padre/madre con trastorno mental severo (TMS)

OpenAIRE

Vivanco B, Gabriela; Grandón F, Pamela

2016-01-01

Introducción. La experiencia de vivir con personas que presentan un Trastorno Mental Severo (TMS) es difícil para las familias, en especial para los hijos quienes han sido poco estudiados. El objetivo de la investigación fue conocer cómo la experiencia de haber vivido con un padre o madre con un trastorno mental severo influyó en la infancia, adolescencia y adultez joven de sus hijos e hijas. Método. Se analizan las experiencias de convivencia con un padre/madre con TMS en 10 hijos (6 hombres...

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.)

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

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-04-15

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

Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

2016-09-15

This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

Aspectos Psicosociales Relacionados con el TEPT en Pacientes con Cáncer de Mama

OpenAIRE

Hernández Moreno, Fresia Paloma; Landero Hernández, René

2015-01-01

El objetivo del estudio fue analizar la relación del estrés, la depresión y las estrategias de afrontamiento con el Trastorno de Estrés Postraumático (TEPT) en una muestra de pacientes con cáncer de mama. Se realizó con una muestra no probabilística de 52 mujeres con cáncer de mama. Se encontró una relación entre TEPT y estrés (r= .344, p= .014), TEPT y depresión (rs=.346, p= .013). El TEPT correlacionó con evitación cognitiva (r= .437, p= .001), TEPT y desesperanza (rs= .437, p= .001) y TEPT...

Análisis de la sensibilidad paramétrica en reactores de lecho fijo

Directory of Open Access Journals (Sweden)

Hermes A. Rangel Jara

1992-05-01

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

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)

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.)

Paciente con esquizofrenia tratado con ziprasidona + clozapina

Directory of Open Access Journals (Sweden)

Pol Yanguas E.

2013-05-01

Full Text Available P es un paciente diagnosticado de esquizofrenia, sigue en un piso tutelado un programa de rehabilitación, está medicado con clozapina 500 mg/día y ziprasidona 280 mg/ día. Padece hipercolesterolemia, tabaquismo y sus hábitos alimenticios no son buenos. La medicación que utiliza desde 2007 hasta ahora se refleja en la tabla 1. El último tratamiento se le introdujo el 7 de agosto de 2012, habiendo presentado un electro cardiograma (ECG normal, pero con ligera taquicardia ventricular y prolactinemia de 44,8 ng/ml (valores normales: 2-18 ng/ml.

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.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

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

REMOVAL OF ORGANIC MATTER AND TOXICITY IN AN UPFLOW IMMOBILIZED BIOMASS ANAEROBIC REACTOR TREATING HOSPITAL WASTEWATER: PRELIMINARY EVALUATION

Directory of Open Access Journals (Sweden)

MÓNICA PORRAS TORRES

2013-01-01

Full Text Available El objetivo de esta investigación consistió en evaluar el desempeño de un reactor anaerobio de flujo ascendente de biomasa inmovilizada (RAFABI tratando un efluente hospitalario real. Se estudió la remoción de materia orgánica y toxicidad, por medio de análisis como UV254, DQOfiltrada y determinación del porcentaje de inhibición en el crecimiento de la raíz de la cebolla. Los resultados mostraron que el proceso biológico estuvo estable durante los 287 días de operación continua, el valor medio de la relación AI/AP fue de 1.21±0.08, indicando que no hubo acumulación de ácidos en el sistema. Sin embargo, los valores de la eficiencia de remoción de DQOfiltrada, 56±15% y UV254, 21±36%, no fueron representativos. La toxicidad se redujo en 50%. Con base en lo anterior, es necesario utilizar el reactor anaerobio en combinación con otros procesos como por ejemplo los procesos de oxidación avanzada, para continuar reduciendo la materia orgánica recalcitrante al proceso anaerobio. Se comprobó la capacidad que tienen los reactores anaerobios de biomasa inmovilizada para remover la toxicidad.

Riesgo familiar total en familias con mujeres diagnosticadas con neoplasia de mama

Directory of Open Access Journals (Sweden)

Nancy Cecilia Charrys-Bravo

2017-07-01

Full Text Available Objetivo: Determinar el riesgo familiar total de las familias con mujeres diagnosticadas con cáncer de mama, que asisten a un centro de salud oncológico en la ciudad de Barranquilla. Materiales y Métodos: El estudio fue de tipo descriptivo, transversal con abordaje cuantitativo. Se utilizó el instrumento Riesgo Familiar Total RFT 5-33. El universo estuvo conformado por familias con mujeres diagnosticadas con cáncer en mama. La muestra fue de 41 familias que integraron a 154 miembros; se abordó el estudio de manera censal, y no por medio de una muestra, porque el acceso a la información de la totalidad de las familia fue viable. Resultados: Los resultados mostraron que las familias, en su mayoría, son de tipo 2. El 68 % de las pacientes categorizan sus familias como amenazadas, el 5 % como familias de alto riesgo y un 27 % de las familias con un bajo riesgo. Conclusiones: Los hallazgos encontrados en esta investigación son importantes para las familias, lo cual permitirá establecer acciones y actividades que logren orientar e implementar procesos de atención específicos con el propósito de cuidar a las familias para que se mantengan sanos en un nivel de bajo riesgo; además, desarrollar controles y seguimiento a aquellas familias que se encuentran en un riesgo alto de amenazas, mediante acciones de promoción y prevención de la enfermedad de una manera amplia. Por lo anterior, se deben emprender programas más agresivos de prevención y promoción, especialmente con las familias que asisten en busca de apoyo médico para este padecimiento; de esta forma, se podrán diagnosticar los casos de forma temprana y proceder al respectivo tratamiento.

REDUCCIÓN DE ÓXIDOSDE NITROGENO CON CATALIZADORES ZEOLÍTICOS INTERCAMBIADOS CON COBALTO

Directory of Open Access Journals (Sweden)

Enrique Alexis García Moya

2012-10-01

Full Text Available Se presenta un estudio sobre la reducción catalítica selectiva (RCS de NOx con metano en presencia de oxígeno en exceso con varios catalizadores a base de las zeolitas mordenita, ferrierita y ZSM-5 intercambiadas con diferentes cargas de cobalto y con las correspondientes zeolitas en forma acida. Cuando la mezcla reactiva contenía mayormente NO2 en lugar de NO, los catalizadores ácidos mostraron las más altas velocidades de formación de N2 en condiciones secas. Las mayores actividades se obtuvieron con los catalizadores Co-mordenita, siguiendo en orden de actividad los catalizadores Co-Ferrierita y Co-ZSM-5. El catalizador Co-Mordenita más activo se ensayó con una mezcla reactiva donde predominó el NO en lugar del NO2 bajo condiciones secas e hidrotérmicas y en presencia de SO2. Con la adición de 8% de agua a la mezcla de reacción se observó desactivación reversible, especialmente a bajas temperaturas. La adición de 60 ppm de SO2 disminuyó la velocidad de reacción aproximadamente a la mitad debido posiblemente al envenenamiento de algunos sitios activos.

Producción de hidrógeno a partir del tratamiento anaerobio de vinazas en un reactor UASB

Directory of Open Access Journals (Sweden)

César González-Ugalde

2014-09-01

Bajo condiciones mesofílicas (37 °C, un pH de operación de aproximadamente 5,50, una concentración del sustrato de 20 000 mg DQO/L y un tiempo de retención hidráulica (TRH de seis horas, la producción promedio de hidrógeno obtenida en el reactor UASB fue de 1,68 mL H2/h/L, con una tasa máxima de 13,4 mL H2/h/L. El porcentaje de remoción de DQO en el proceso de fermentación alcanzó valores máximos del 43%, con un promedio cercano al 20%. Tanto la producción de hidrógeno como la remoción de DQO presentaron una dependencia inversamente proporcional al TRH. Los resultados obtenidos en este estudio demuestran que la fermentación anaerobia en un reactor UASB abre la posibilidad de utilizar las vinazas para producir hidrógeno molecular de forma sostenible.

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

International Nuclear Information System (INIS)

Anon.

1986-01-01

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

EVALUACIÓN DE LA SENSIBILIDAD PARAMÉTRICA DEL PROCESO DE SÍNTESIS DE LA CICLONITA EN UN REACTOR POR LOTES

Directory of Open Access Journals (Sweden)

Juan Carlos Ojeda Toro

Full Text Available En este trabajo se evaluó la sensibilidad paramétrica del proceso de síntesis de la ciclonita en un reactor por lotes. Esto con el fin de definir condiciones seguras de operación para esta reacción altamente exotérmica. La ley de velocidad de la reacción se ajustó a partir de datos experimentales disponibles en la literatura. Reparametrizando los balances de materia y energía del reactor, se estableció la sensibilidad de la temperatura de reacción con respecto a la variación de la temperatura inicial del sistema reactivo y la temperatura del medio refrigerante. Para determinar las condiciones críticas de operación del reactor, se usó como criterio el cálculo de los coeficientes de sensibilidad y los perfiles de temperatura-conversión así como el lugar geométrico de los máximos de estas curvas. Se definió para el sistema reactivo un potencial crítico de generación de calor (M igual a 34 y que las condiciones críticas de Runaway corresponden a un número de Semenov (ψ igual a 0,684, un parámetro de calor de reacción (B igual a 15 y un número del tipo Arrhenius (γ con un valor de 20. Así mismo, los perfiles de temperatura-conversión precisan una relación crítica entre el potencial de enfriamiento y generación de calor de 2,5786 (N/M.

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

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

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.)

Terapias alternativas con animales para niños con necesidades especiales

OpenAIRE

Cea-Chueca, Aihnoa

2014-01-01

El presente trabajo de fin de grado consiste en la realización de un análisis y descripción de la literatura acerca de las distintas terapias animales, destinadas a niños con necesidades especiales: cómo y por qué surgieron, en qué consisten las terapias, cómo podemos utilizarlas, etc. Las terapias alternativas con animales se conocen como una alternativa terapéutica, donde se utiliza a un animal como co-terapeuta para poder desarrollar diferentes capacidades de los niños con necesidades espe...

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)

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

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

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

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)

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

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

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

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

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.)

Variables socionutricionales de hogares mazahuas integrados por preescolares desnutridos con madres con obesidad y sin obesidad

Directory of Open Access Journals (Sweden)

Conzuelo González, Viridiana Vanessa

2009-01-01

Full Text Available El primer objetivo fue conocer cuántos menores de cinco años con diferentes grados de desnutrición tienen una madre con sobrepeso/obesidad/ en una comunidad indígena que vive en extrema pobreza y bajo condiciones de migración masculina internacional. El segundo fue comparar tres variables socionutricionales (ingreso familiar, educación de la madre y adecuación nutrimental de la dieta diaria entre estos hogares y los hogares con desnutrición infantil y madres sin obesidad. Se realizó un estudio transversal (2006-2007, en la comunidad mazahua de San Francisco Tepeolulco, Municipio de Temascalcingo; que incluyó a 85 hogares integrados por preescolares con desnutrición inscritos al programa Oportunidades. Se determinó el estado nutrición de los preescolares con indicadores antropométricos y se obtuvo el IMC de las madres de estos infantes. Se aplicó una encuesta socionutricional, incluida el recordatorio de 24 horas, y complementado con la observación participante (cualitativa. Se encontró que 83% de las madres mazahuas presentaron sobrepeso u obesidad. El estado de nutrición de los preescolares con madres con obesidad presentó un porcentaje mayor de desnutrición (76%. En la variable género, se encontró que 54% de los niños con madres con obesidad tenía baja talla. Al relacionar el nivel educativo de la madre, esta variable resultó ser estadísticamente significativa (p=0.015, donde el analfabetismo está más relacionado con la desnutrición infantil que tienen madres de bajo y/o peso normal. La elevada prevalencia de hogares conformados con preescolares con desnutrición y madres con obesidad, es un síntoma más de la pobreza en zonas indígenas en México, con bajo índice de desarrollo humano.

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

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

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

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.

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

Evaluación del comportamiento hidráulico en un reactor anaerobio de doble cámara (RADCA

Directory of Open Access Journals (Sweden)

Nancy Rincón

2011-01-01

tales como cortos circuitos, zonas muertas y recirculación interna afectan su desempeño. En esta investigación se evaluó el comportamiento hidráulico de un reactor anaerobio de doble cámara (RADCA de 534,5 L (cámara 1=305 L y cámara 2= 229,5 L como innovación tecnológica de los reactores UASB. El RADCA fue alimentado con agua residual municipal (ARM de la ciudad de Maracaibo, Venezuela; cada una de las cámaras fueron inoculadas con lodo granular (20% v/v proveniente de una cervecería local. La evaluación hidráulica se realizó en la fase líquida y en operación utilizando Li+ (LiCl como trazador aplicado de forma instantánea en el afluente a tiempo de retención hidráulico teórico (TRHt de 6 horas; 3,4 h en la cámara 1 y 2,6 para la cámara 2. El RADCA describió un flujo pistón en ambas cámaras y una eficiencia hidráulica cercana a la unidad (1 indicando una presencia casi nula de zonas muertas. La eficiencia de remoción de la DQO total (DQOT del RADCA se mantuvo en el rango de 59,77% a 74,64% con un promedio de 68,26%. Para las cámaras 1 y 2 la eficiencia promedio fue 60,4 y 20,94% con una producción de biogás (L/h de 2,768 y 0,541 respectivamente.

Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

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.)

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.

DEGRADACIÓN ELECTROQUÍMICA DE FENOL: SIMULACIÓN NUMÉRICA Y CONTROL CON ESTRUCTURAS BÁSICAS DE CONTROL

Directory of Open Access Journals (Sweden)

Alejandro Regalado-Méndez

2012-01-01

Full Text Available En este trabajo se propone un control PI feedback para mejorar el desempeño de la dinámica de un reactor electroquímico de tanque agitado (RECTA que degrada fenol. Esto con la finalidad de mejorar la calidad de aguas residuales de acuerdo a la normatividad mexicana (CDQO = 200 mg L-1 por día. La ley de control tiene como variable de control a la concentración de DQO, mientras que la variable manipulada es la tasa de disolución. El principal resultado es que la ley de control implementada es robusta al llegar a la referencia pedida(CDQO, set-point = 83.92 mg L-1. Es decir, regula perfectamente la concentración de DQO, sin embargo el tiempo de asentamiento del reactor es mayor que a lazo abierto.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Terapia regenerativa con plasma rico en plaquetas en pacientes con quemaduras

Directory of Open Access Journals (Sweden)

María del Carmen Franco Mora

2015-10-01

Full Text Available Se realizó un estudio comparativo, longitudinal y prospectivo de 60 adultos con quemaduras profundas, atendidos en el Hospital General docente "Dr. Juan Bruno Zayas Alfonso" de Santiago de Cuba, desde febrero de 2013 hasta igual mes de 2014, para evaluar la efectividad de la terapia regenerativa con plasma rico en plaquetas para la cicatrización de estas lesiones. La muestra fue dividida en 2 grupos de 30 integrantes cada uno: a los del primero se les aplicó plasma rico en plaquetas y a los del segundo sulfadiazina de plata. Se utilizaron las frecuencias absolutas y el porcentaje como medidas de resumen para variables cualitativos, así como la media aritmética y la desviación estándar para las cuantitativas. Con el proceder aplicado la cicatrización se completó en un tiempo menor, de manera que se demostró la efectividad de esta alternativa terapéutica en relación con el tratamiento convencional

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

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium; Diseno de un reactor nuclear subcritico heterogeneo con sales fundidas a base de torio

Energy Technology Data Exchange (ETDEWEB)

Medina C, D.; Hernandez A, P.; Letechipia de L, C.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080-A (Venezuela, Bolivarian Republic of)

2015-09-15

This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a {sup 252}Cf source, whose dose levels at the periphery allows its use in teaching and research activities. The design was realized by the Monte Carlo method, where the geometry, dimensions and the fuel was varied in order to obtain the best design. The result was a cubic reactor of 110 cm of side, with graphite moderator and reflector. In the central part having 9 ducts of 3 cm in diameter, eight of them are 110 cm long, which were placed on the Y axis; the separation between each duct is 10 cm. The central duct has 60 cm in length and this contains the {sup 252}Cf source, also there are two irradiation channels and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For the design the k{sub eff} was calculated, neutron spectra and ambient dose equivalent. In the first instance the above was calculated for a virgin fuel, was called case 1; then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose of comparing two different fuels operating within the reactor. For the two irradiation ducts three positions are used: center, back and front, in each duct in order to have different flows. (Author)

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

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

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

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

RadConEd: A Graphical Data Editor for the Radiological Consequences Model, RadCon

International Nuclear Information System (INIS)

Crawford, J.; Domel, R.U.

2000-05-01

This document describes the application, RadConEd, which has been designed and implemented to enable users of the RadCon system to update these parameter files. The RadCon system is written in the Java programming language, and as such provides portability across computer platforms. The software described in this report was developed in line with the portability requirements of RadCon, thus providing a uniform user interface across computer platforms and bypassing the need of using system editors. In addition a number of data integrity measures were implemented

Experimentos de combustión con cascarilla de arroz en lecho fluidizado para la producción de ceniza rica en sílice

Directory of Open Access Journals (Sweden)

Juan Daniel Martínez Ángel

2010-01-01

Full Text Available A fin de contribuir con la valorización de residuos agrícolas que presentan dificultades de disposición, la utilización de energéticos renovables y la obtención de formas de energía más limpias y amigables con el medio ambiente, se investigó la combustión de cascarilla de arroz en un reactor de lecho fluidizado burbujeante y atmosférico. La metodología consistió en evaluar la influencia de altos excesos de aire, en la obtención de gases calientes y ceniza con altos contenidos de sílice en estado amorfo. Excesos de aire en el rango de 40% al 125%, mostraron que los valores cercanos al límite inferior establecido (40% favorecieron temperaturas superiores a 700ºC en el interior del reactor, dando origen a mayores eficiencias de conversión de la cascarilla, pero disminuyendo el potencial amorfo de la sílice contenida en la ceniza. Comportamiento contrario fue evidenciado en el límite superior del exceso de aire. Las concentraciones de los gases del proceso mostraron relativa variabilidad para cada condición de operación. Para el caso del CO, se reportaron emisiones promedio de 200 ppm (en una base seca del 11% de O2, para los excesos de aire comprendidos entre 40% y 82,5%.

Optimization of the biological process using flat membrane bioreactors. Maximum treatment performance with minimum reactor volume; Optimizacion del proceso biologico con BRM de membrana plana. Maximo rendimiento de depuracion con minimo volumen de reactor

Energy Technology Data Exchange (ETDEWEB)

Lluch Vallmithana, S.; Lopez Gavin, A.

2006-07-01

In a conventional activated sludge process, the membranes are inside the biological reactor where they drain the water through suction or a water column. This system can be operated with heavy loads and sludge of 12-14 g/l or more, and is not affected by problems of bulking or foaming. This makes it suitable for treating difficult industrial waste waters, providing treated water that is free of bacteria and viruses. Micro filtration membranes are flat without any rubbing between them. The membranes require infrequent chemical cleaning and do not need back washing. As no final sedimented is needed, the waste water treatment plant occupies less space. (Author)

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Calidad de vida, satisfacción con el tratamiento y bienestar emocional en pacientes con diabetes LADA

OpenAIRE

Granado Casas, Minerva

2014-01-01

Introducción: La calidad de vida, satisfacción con el tratamiento y bienestar emocional en pacientes con diabetes mellitus están relacionados con el control metabólico, las complicaciones y algunas variables socioeconómicas. Actualmente, no existe ningún estudio científico que analice la calidad de vida, satisfacción con el tratamiento y bienestar emocional en pacientes con diabetes LADA. Objetivos: Los objetivos principales fueron conocer la calidad de vida, satisfacción con el tratamient...

Core reactor simulation of the Central Laguna Verde (CLV) reactor in stationary state and an example of the application in the recharge options analysis of cycle 3; Simulacion del nucleo del reactor de la Central Laguna Verde (CLV) en estado estacionario y ejemplo de aplicacion en el analisis de alternativas de recarga del ciclo 3

Energy Technology Data Exchange (ETDEWEB)

Ocampo Mansilla, Hector; Francois Lacouture, Juan Luis; Blanco Lara, Jesus; Cortes Campos, Carlos Cristobal; Esquivias Montoya, Jesus; Esquivel Torres, Jose Luis; Martin del Campo Marquez, Cecilia [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico); Sanchez Herrera, Luciano; Torres Alvarez, Carlos [Comision Federal de Electricidad (CFE), Mexico, D. F. (Mexico)

1991-12-31

The results are presented of a study requested by Comision Federal de Electricidad (CFE) for the analysis of Cycle 3, of Unit No. 1 of the Laguna Verde Nuclear Power Station (CLNV) and determine the burning effect impact, carried out with the starting tests and the operation of Cycles 1 and 2 on base of the cycle extension known as coastdown. The calculations were realized with the Code Package FMS for fuel managing, using the Code PRESTO-B that analyzes the reactor in detailed form in three dimensions an in stationary state. In the study the schemes of fraction of recharge proposed by General Electric (GE) were analyzed with the effect of cycle extension. The initial design value of 100 assemblies for Cycle 3, GE proposes to increase such fraction from 112 to 120 assemblies. This impacts the cost of the second recharge and the purpose of this investigation is to analyze options with higher fuel enrichment in U-235 to minimize the number of assemblies in this recharge. The analyses effected show that the designs proposed by GE do not fulfill the required energy proposed for the cycle, even using in the recharge only fuel with 3.03% of enrichment. It is proposed, likewise, the fuel enrichment up to 3.25% to satisfy the energy demand with a minimum of assemblies. [Espanol] Se presentan los resultados de un estudio solicitado por la Comision Federal de Electricidad (CFE) para analizar el ciclo 3, de la unidad 1 de la Central Laguna Verde (CLV), y determinar el impacto del efecto de quemado llevado a cabo con las pruebas de arranque y por la operacion de los ciclos 1 y 2 con base en la tecnica de alargamiento del ciclo conocida como coastdown1. Los calculos se realizaron con el paquete de codigos FMS para la administracion de combustible, usando el codigo PRESTO-B que analiza el reactor en forma detallada en tres dimensiones y en estado estacionario. Se analizaron en el estudio los esquemas de fraccion de recarga propuesta por la General Electric (GE) con el efecto de

Core reactor simulation of the Central Laguna Verde (CLV) reactor in stationary state and an example of the application in the recharge options analysis of cycle 3; Simulacion del nucleo del reactor de la Central Laguna Verde (CLV) en estado estacionario y ejemplo de aplicacion en el analisis de alternativas de recarga del ciclo 3

Energy Technology Data Exchange (ETDEWEB)

Ocampo Mansilla, Hector; Francois Lacouture, Juan Luis; Blanco Lara, Jesus; Cortes Campos, Carlos Cristobal; Esquivias Montoya, Jesus; Esquivel Torres, Jose Luis; Martin del Campo Marquez, Cecilia [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico); Sanchez Herrera, Luciano; Torres Alvarez, Carlos [Comision Federal de Electricidad (CFE), Mexico, D. F. (Mexico)

1992-12-31

The results are presented of a study requested by Comision Federal de Electricidad (CFE) for the analysis of Cycle 3, of Unit No. 1 of the Laguna Verde Nuclear Power Station (CLNV) and determine the burning effect impact, carried out with the starting tests and the operation of Cycles 1 and 2 on base of the cycle extension known as coastdown. The calculations were realized with the Code Package FMS for fuel managing, using the Code PRESTO-B that analyzes the reactor in detailed form in three dimensions an in stationary state. In the study the schemes of fraction of recharge proposed by General Electric (GE) were analyzed with the effect of cycle extension. The initial design value of 100 assemblies for Cycle 3, GE proposes to increase such fraction from 112 to 120 assemblies. This impacts the cost of the second recharge and the purpose of this investigation is to analyze options with higher fuel enrichment in U-235 to minimize the number of assemblies in this recharge. The analyses effected show that the designs proposed by GE do not fulfill the required energy proposed for the cycle, even using in the recharge only fuel with 3.03% of enrichment. It is proposed, likewise, the fuel enrichment up to 3.25% to satisfy the energy demand with a minimum of assemblies. [Espanol] Se presentan los resultados de un estudio solicitado por la Comision Federal de Electricidad (CFE) para analizar el ciclo 3, de la unidad 1 de la Central Laguna Verde (CLV), y determinar el impacto del efecto de quemado llevado a cabo con las pruebas de arranque y por la operacion de los ciclos 1 y 2 con base en la tecnica de alargamiento del ciclo conocida como coastdown1. Los calculos se realizaron con el paquete de codigos FMS para la administracion de combustible, usando el codigo PRESTO-B que analiza el reactor en forma detallada en tres dimensiones y en estado estacionario. Se analizaron en el estudio los esquemas de fraccion de recarga propuesta por la General Electric (GE) con el efecto de

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

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Desalination plan with nuclear reactors as part of a sustainable development program in Mexico; Plan de desalinizacion con reactores nucleares como parte de un programa de desarrollo sustentable en Mexico

Energy Technology Data Exchange (ETDEWEB)

Rojas A, O; Calleros M, G., E-mail: oziel.rojas.siimisa@gmail.com [Soluciones en Instrumentacion Integral y Mantenimiento Industrial y Servicios, S. A. de C. V. (Mexico)

2016-09-15

This paper presents a project for the desalination of seawater with nuclear reactors, in order to supply fresh water to the populations near to the nuclear power plant. A case study is proposed with the nuclear power plant of Laguna Verde, implementing a system that allows taking advantage of the residual heat of the seawater condensate stage and with this, to supply drinking water to the surrounding localities where the vital liquid is scarce. In addition, legislation is proposed to allow some of the thermal energy generated by reactors producing electrical energy in Mexico to be used for the desalination of seawater and/or hydrogen production. (Author)

Continuous aerobic treatment of dairy waste waters. Supported reactor; Depuracion aerobia de aguas residuales de industrias lacteas en regimen continuo. Reactor con soporte

Energy Technology Data Exchange (ETDEWEB)

Carta, F.; Romeero, F.; Pereda, J.; Alvarez, P. [Universidad de Sevilla (Spain)

1998-12-31

Experiences in continuous flow been carried out to achieve effluents with low COD and a minimum ammonium nitrogen concentration. A 80 liter reactor with rectangular cross-section (width 25 cm and depth 30 cm) and length 150 cm was used thermostated. The system was aerated by air injection the air passed through perforated tubes settled at the bottom and covered by a plastic mesh. The liquor (water and milk) temperature was near 30 degree centigree Mixtures of milk and water, with pH of 11 and containing 3500 mg COD/1 (as sewage from Dairy Center`s) were fed into the reactor. A mixed cultive constituted of cultives isolated from Dairy Center`s effluent adapted to 30 degree centigree for 19 days, and mixed with sludge from a domestic wastewater treatment plant was used as inoculum. The influent flow rates were 81/d, 101/d and 12,6 L/d. The experiences went on up to stationary-state. In all the experience, the pH values become stabilized over 8,5 and the nitrite and nitrate nitrogen concentrations were insignificant. It was observed that when flow rate rises, the average COD values and the ammonium nitrogen concentrations (achieved at the end of experiences) decreased in a parallel way up to a certain flow rate value at which they are almost constant with the increase of the flow rate. The sludge analysis showed a composition of 5,4 gN and 2,4 gP in 100 g biomass. (Author) 16 refs.

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.)

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

Efecto de la concentración de SO2 en las reacciones de calcinación y sulfatación de calcáreos en reactores de lecho fluidizado. // Effect of the SO2 concentration in the calcinations and sulfatation reactions in a fluidized bed reactor.

Directory of Open Access Journals (Sweden)

J. E. Lindo Samaniego

2008-01-01

Full Text Available Fue realizado un estudio sobre el efecto de las concentraciones de SO2 en la absorción por calcáreo en hornos de LechoFluidizado. Para observar la influencia del SO2 en los parámetros de diferentes procesos físicos y químicos fueron creadosambientes para cuatro concentraciones diferentes de SO2 : 500, 1000, 2000 y 4000 ppm. Se utilizaron dos tipos decalcáreos: Dolimitico-DP y el Calcítico-CI. El Lecho Fluidizado Burbujeante utilizado tiene 160 mm de diámetro interno yfue fluidizado con aire a la temperatura de 850 °C, con una concentración de SO2 deseada. Como material del lecho fueutilizada la arena de cuarzo (99,9% con diámetro de 385 μm y de masa aproximadamente 3,0 kg. El calcáreo fueadicionado en dosificación de 50 g con el reactor ya pre-calentado. Las variaciones de las concentraciones de SO2, CO2,CO, O2 y las descargas, fueron monitoreadas continuamente a la salida del ciclón que fue utilizado para la retención de lapartícula fina. Para esos dados se desarrolló un programa en LabView. El modelo matemático escogido posibilitó ladeterminación de la conversión.Palabras claves: Lecho fluidizado, dióxido de azufre, absorción de azufre, calcáreo , reactor de lechofluidizado.___________________________________________________________________________Abstract.A Study of the effect of the concentrations of SO2 in its absorption by limestones in fluidized bed furnaces wasconducted. For the determination of the SO2 influence on the different physical and chemical parameters of process,such as calcinations and sulfatation four different atmospheres were used in the reator with concentrations of SO2 of500, 1000, 2000 and 4000 ppm. Two types of limestones were used: Dolomite-DP and Calcitic-CI. The bench scalebubbling fluidized bed reactor had a 160 mm internal diameter and was fluidized with air at 850 °C containing therequired concentration of SO2. Bed material was quartz sand (99,9% , with 385 μm diameter and approximately

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

Directory of Open Access Journals (Sweden)

Carlos Ramiro Escalera Vásquez

2005-01-01

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

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.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

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.)

Mastopexia con prótesis: técnica triplanar con colgajo en cola de pez para mamas con pobre calidad de cobertura

Directory of Open Access Journals (Sweden)

M. Berrocal-Revueltas

2013-12-01

Full Text Available La mastopexia con implantes de silicona, es uno de los procedimientos más frecuentes en la práctica diaria en Cirugía Plástica. Sin embargo, la pobre calidad de la cobertura de las mamas, relacionada con piel estriada, escaso tejido subcutáneo y pérdida del parénquima mamario secundarios a la lactancia prolongada y a la pérdida masiva de peso, son causa del incremento progresivo de resultados insatisfactorios a medio y largo plazo que han sido evaluados, analizados y discutidos por los especialistas. Los resultados insatisfactorios más frecuentes en mastopexia con implantes en posición retromuscular son: doble contorno mamario, asimetrías y desplazamiento de los implantes. Los objetivos del presente estudio son: describir el nuevo método triplanar como alternativa útil para lograr mejores resultados en pacientes con mala calidad de cobertura y analizar los resultados obtenidos con el mismo. Realizamos un estudio observacional, descriptivo, prospectivo, en un grupo de 268 pacientes con diagnóstico de ptosis mamaria con hipotrofia y mala calidad de cobertura por piel estriada, escaso tejido celular subcutáneo y pérdida del parénquima mamario, secundarios a lactancia prolongada y pérdida masiva de peso, entre otros, a las que se les practicó mamoplastia de aumento con implantes de silicona mediante la aplicación del método triplanar descrito por la autora, entre enero del 2004 y enero del 2013 en el Hospital de Bocagrande de Cartagena, Colombia. El diseño de la técnica se basa en los principios de cicatriz vertical descrita por Lassus o Lejour o mediante cicatriz en "J", descrita por la autora. El método triplanar se basa en tres planos de cobertura del implante. Plano I: submuscular, para cubrir los dos tercios superiores del implante. Plano II: colgajo dermoglandular con dos prolongaciones distales triangulares de tejido adiposo, en forma de pez, para cubrir el tercio inferior del implante y el borde libre del m

Inducción fúngica de la biorremediación de suelos contaminados con combustibles

Directory of Open Access Journals (Sweden)

Mary Lopretti

2011-04-01

Full Text Available La biorremediación de suelos ha sido en los últimos años una de las aplicaciones de los procesos de la biotecnología industrial que se ha desarrollado en busca de soluciones  naturales y eficientes al problema de la contaminación.Varios son los microorganismos que viven en condiciones extremas, desarrollando estrategias metabólicas que les permiten tener sus vías metabólicas aptas para vivir y reproducirse. Dentro de estas estrategias  se encuentra la acción de enzimas hidrolíticas, oxidativas y depolimerizantes que permiten modificar sustratos complejos como lo son los derivados del petróleo  y entre ellos los combustibles.En el presente trabajo se estudió la acción de dos hongos Gloeophylum trabeum y Phanerochaetes chrysosporium actuando en forma consorcial sobre sustratos  contaminados con gasoil y nafta.Se prepararon 4 reactores de fermentación sólida: dos de ellos blanco, contaminados sin inducción, y de los otros dos uno con nafta y el otro con gasoil inoculados ambos con 100cc de medio de propagación con micelios de P.chrysosporium y 50cc de medio de crecimiento con pellets de G. trabeum. Se extrajeron muestras cada 15 días y se evaluó  en el material de cada reactor características fisicoquímicas como %C, humedad y pH.También se realizó la evaluación microbiológica por siembra en placa y dilución en placa con agar malta al 1,25%. En todos los casos se obtuvieron micelios.Por último se determinó la actividad de enzimas lacasa  y peroxidasa. La actividad lacasa se determinó por espectrometría usando 0.5mM de ABTS como sustrato en 0,1M de buffer acetato de sodio pH5 y 1 ml de extracto obtenido por extracción salina. La determinación de actividad Mn peroxidasa se realizó con 0.01% de rojo fenol como sustrato en buffer succinato de sodio 0.1M en presencia de 1 ml de extracto enzimático, MnSO4 0.1M y H2O2 0.1M.Las actividad  de éstas enzimas permitió obtener moléculas mas pequeñas producto de la

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.

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

Estudio de la recuperación de cromo hexavalente mediante un reactor electroquímico de compartimentos separados por separadores cerámicos

OpenAIRE

REYES PINEDA, HENRY

2011-01-01

La Tesis Doctoral "Estudio de la recuperación de cromo hexavalente mediante un reactor electroquímico de compartimentos separados por separadores cerámicos" se centra en la posibilidad de recuperación del cromo hexavalente procedente de las disoluciones de mordentado de las industrias de metalizado de plásticos mediante la utilización de un reactor electroquímico de compartimentos separados por separadores cerámicos fabricados a diferente presión y diferente composición de almidón. Con la rec...

Estudio comparativo del efecto del cepillado con una crema dental con propóleos rojos y de un gel con clorofila

OpenAIRE

Estela Gispert Abreu; Elena Cantillo Estrada; Aracelys Rivero López; Berta Oramas Rodríguez

1998-01-01

Se analiza comparativamente el efecto de un gel dental de clorofila y de una crema dental con propóleos rojos sobre varios parámetros relacionados con la caries dental, en escolares que se cepillaron durante 21 días con dichos productos. Se obtuvieron resultados favorables principalmente en la disminución del grado de infección por Streptococcus mutans y la elevación de la capacidad individual de remineralización; salvo en este último no hubo diferencias estadísticamente significativas.Author...

Validation of the inspections with ultrasound of the welds of the reactor of ITER vacuum vessel; Validacion de las inspecciones con ultrasonidos de las soldaduras de la Vasija de Vacio del reactor del ITER

Energy Technology Data Exchange (ETDEWEB)

Garcia, A.; Fernandez, F.; Perez, C.; Sillero, J. A.

2013-07-01

The ITER fusion reactor vacuum vessel has thousands of welding austenitic with shapes and different manufacturing processes. The RCC-MR code, which is that applied to the manufacture of the fusion reactor, requires a volumetric test all of them. This test should be mainly by x-rays and welds where it was not possible to use this method, ultrasonic.09-06.

Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

Energy Technology Data Exchange (ETDEWEB)

Campos Gonzalez, Rina Margarita

2007-07-15

developed work only concentrates in the reactor core, but taking advantage of the modularity that PVM offers, it is possible to add component such as separators and steam dryers, lines of steam and feed water to obtain a model of a complete closed circuit. The applications concentrate mainly in the training of personnel in the phenomenology of the BWR, and as an investigation tool in the study of the dynamics of BWR reactors. The oscillations out of phase study presents challenges at the moment as are the explanation of the variation of the neutral line with time, non azimuthal but axial oscillations out of phase, etc. So far a first model oriented in this direction is at hand. [Spanish] Este trabajo consiste en la integracion de tres modelos desarrollados previamente los cuales se encuentran ampliamente descritos en la literatura: modelo del canal termohidraulico, modelo de la neutronica modal y el modelo de los lazos de recirculacion. La herramienta utilizada para este acoplamiento de modelos es el sistema PVM, Parallel Virtual Machine, que permitio paralelizar el modelo mediante el concepto de computacion distribuida. La finalidad de hacer este acoplamiento de modelos es la de obtener una herramienta mas completa que represente mejor la configuracion real y la fenomenologia del nucleo de un reactor BWR, obteniendo asi mejores resultados. Ademas mantener la flexibilidad de mejorar el modelo resultante en cualquier momento, ya que los modelos muy complejos o sofisticados resultan dificiles de mejorar siendo imposible modificar las ecuaciones que utilizan y pueden incluir variables que no son de importancia primaria en el problema tratado o que enmascaren relaciones entre variables debido al exceso de resultados. Tambien el mantener la flexibilidad de agregar modelos de componentes o sistemas del reactor BWR, todo esto dependiendo de las necesidades del modelado. Se eligio a la planta sueca Ringhals para caracterizar el modelo acoplado resultante por contar con un Benchmark

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.)

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

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.

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Estudio de distribución de tiempos de residencia en un reactor biológico de lecho empacado cerámico

Directory of Open Access Journals (Sweden)

Tatiana Rodríguez Chaparro

2004-01-01

Full Text Available La distribución de tiempos de residencia de un reactor es una característica del mezclado que ocurre dentro de él [ ] 1 [ ] 2 ; su determinación es básica para el diseño de cualquier tipo de reactor en escala real. El objetivo de esta investigación consistió en determinar la distribución de tiempos de residencia en un reactor biológico de lecho empacado cerámico (anillos a partir de pruebas con trazadores. Los resultados obtenidos utilizando las técnicas de inyección por paso y pulso fueron 34.577 seg., y 17.745 seg., respectivamente, y la dispersión calculada infinita. Lo anterior permite concluir que en reactores de lecho empacado cerámico (anillos las moléculas del trazador se distribuyen uniformemente en todo el sistema. Los ensayos se realizaron en un modelo a escala laboratorio.

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

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

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

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

Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

Energy Technology Data Exchange (ETDEWEB)

Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

2015-09-15

This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

Investigando con personas con dificultades de aprendizaje

Directory of Open Access Journals (Sweden)

Borja González Luna

2013-12-01

Full Text Available El artículo muestra los orígenes de lo que Walmsley (2008 denomina «investigación inclusiva». Para comprender qué se entiende por investigación inclusiva tenemos que remontarnos a los debates epistemológicos sobre las metodologías cuantitativas y cualitativas, acontecidos en la década de los 90, en torno a la revista Disability & Society. A partir de una síntesis de dichos debates, focalizados en el ámbito de la «discapacidad intelectual y del desarrollo», se exponen dos estrategias de colaboración con dicha población: a una aproximación etnográfica (de trabajo grupal, y b una aproximación biográfica (de trabajo individual. A continuación se esboza un posible diseño de trabajo de campo que intenta superar el paradigma cualitativo «clásico» con el objetivo de incluir a dicho colectivo más allá del rol de «sujetos de la investigación». Para finalizar se recoge el debate sobre la accesibilidad de los resultados de la investigación a los participantes en dichas investigaciones, y con ello la necesaria innovación en el ámbito de las «devoluciones» de los resultados, cuando se trata de incluir a personas que presentan limitaciones para la comprensión del lenguaje abstracto oral y/o escrito.

MODELADO CON REDES DE PETRI E IMPLEMENTACIÓN CON GRAFCET DE UN SISTEMA DE MANUFACTURA FLEXIBLE CON PROCESOS CONCURRENTES Y RECURSOS COMPARTIDOS

OpenAIRE

Johanna Stella Castellanos Arias; Leonardo Enrique Solaque Guzmán

2010-01-01

En este trabajo, se presenta el modelado de un Sistema de Manufactura Flexible (SMF), con procesos concurrentes y recursos compartidos mediante Sistemas a Eventos Discretos (SED), específicamente Redes de Petri (RdP), y GRAFCET. El SMF se plantea como un modelo hipotético que se modela con una RdP con el objeto de identificar su dinámica y hallar la secuencia óptima de funcionamiento del sistema. Se desarrolló un modelo matemático que permite estimar el vector de tiempo acumulado de un proces...

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

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...

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.)

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

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

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

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)

Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety

International Nuclear Information System (INIS)

Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.

1989-01-01

A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

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

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

con mala calidad de vida

Directory of Open Access Journals (Sweden)

Agustín Martín-Rodríguez

2007-01-01

Full Text Available En este estudio ex post facto se ha analizado si los familiares de pacientes con mala calidad de vida presentan diferencias en las variables clínicas de personalidad y relaciones familiares en función de que el paciente haya estado o no ingresado en una Unidad de Cuidados Intensivos. Seleccionamos dos grupos: 29 familiares de pacientes traumatizados graves transcurridos cuatro años de su ingreso en una UCI de Traumatología y con mala calidad de vida (debido a secuelas físicas y/o psicológicas tras el ingreso, tales como traumatismos craneoencefálicos, politraumatismos y tetraplejias traumáticas y 32 familiares de pacientes con mala calidad de vida con cuatro años de evolución de su enfermedad física (hipertensión, diabetes, artritis reumatoide y síndrome de intestino irritable que no han estado ingresados en la UCI. Para alcanzar nuestro objetivo empleamos una Encuesta Psicosocial y los siguientes instrumentos: Cuestionario de Análisis Clínico, Escala de Clima Social en la Familia y Escala de Adaptación Psicosocial de la Enfermedad. Los resultados mostraron que los familiares de pacientes con mala calidad de vida que estuvieron ingresados en la UCI hace cuatro años, presentan diferencias significativas en las variables agitación y expresividad comparados con los familiares de pacientes con mala calidad de vida que no han estado ingresados en la UCI.

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.)

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

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

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

Directory of Open Access Journals (Sweden)

Caicedo Luis A.

1999-06-01

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

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

TRATAMIENTO DE AGUAS RESIDUALES DE UNA INDUSTRIA PROCESADORA DE PESCADO EN REACTORES ANAERÓBICOS DISCONTINUOS

Directory of Open Access Journals (Sweden)

Julio César Marín Leal

2015-01-01

Full Text Available En el presente trabajo se evaluó el tratamiento de las aguas residuales de una industria procesadorade pescado de la ciudad de Manta (Ecuador, en reactores anaeróbicos discontinuos, y se estableciósu adecuación a las normas ambientales vigentes en materia de vertido. Para ello, se realizaronensayos de laboratorio en reactores discontinuos de 1 L, con un tiempo de contacto de 24 h yprovistos de un lodo anaerobio procedente de una planta de tratamiento de aguas residualesdomésticas. Dicho efluente fue diluido con agua destilada en proporciones de 33%, 66% y 100%,correspondientes a las etapas I, II y III, respectivamente. Durante cada etapa se monitorearon los siguientes parámetros, de acuerdo con los métodos estándares: pH, alcalinidad total, DBO5.20,DQO, nitrito, amonio, nitrógeno total Kjeldahl (NTK, ortofostato, sulfato, sólidos suspendidostotales (SST y sólidos suspendidos volátiles (SSV. Los resultados muestran porcentajes deremoción de materia orgánica, expresados como DBO5.20(37.9±4.1%; 41.8±7.6% y 46.2±3.2% yDQO (34.7±9.7%; 36.9±9.2% y 43.8±4.1%, para las etapas I, II y III, respectivamente, relativamentebajos como resultado del origen del inóculo usado, así como del contenido relativo de sales en elefluente industrial. Las remociones de amonio, NTK y ortofosfato estuvieron entre 60-95%, 25-37% y 6-25%, respectivamente. Bajo las condiciones de los ensayos realizados, el efluente tratadorequiriere de la aplicación de un postratamiento para reducir el contenido de materia orgánica ynutrientes a los límites permisibles de descarga establecidos en la República de Ecuador.

Evaluación de la co-digestión anaerobia de lodos de aguas residuales municipales con residuos de alimentos

Directory of Open Access Journals (Sweden)

Ileana Consuelo Julio Guerrero

2016-06-01

Full Text Available La digestión anaerobia es un proceso ampliamente utilizado para el tratamiento de los lodos producidos en las plantas de tratamiento de agua residual, debido a sus ventajas técnicas y económicas. Este artículo presenta un estudio en el que se evaluó la co-digestión de lodos de aguas residuales con residuos de alimentos (RA como una estrategia para optimizar la digestión de los lodos. Se realizaron mono-digestiones y co-digestiones de los sustratos en condiciones mesofílicas (35°C usando reactores por lotes. Los lodos utilizados fueron: lodo primario (LP, lodo secundario espesado (LSE y una mezcla de LP con LSE en la proporción 60:40 con base en sólidos totales (LP:LSE. Las co-digestiones se realizaron usando distintas proporciones de mezclas de los sustratos con base en sólidos volátiles totales: LP:RA=30:70, LP:RA=50:50, LP:RA=70:30 y (LP+LSE:RA=70:30. La máxima producción de metano, 0,25LCH4/gSVadicionado, se obtuvo para la mezcla LP:RA=30:70, presentando una producción 32% mayor que la obtenida en la mono-digestión del lodo primario.

Calidad de vida relacionada con la salud del niño y del adolescente con obesidad

OpenAIRE

Hurtado-Valenzuela, Jaime Gabriel; Álvarez-Hernández, Gerardo

2014-01-01

Objetivo Estimar la calidad de vida relacionada con la salud (CVRS) en un grupo de niños y adolescentes con obesidad de la ciudad de Hermosillo, Sonora, México. Métodos Se realizó un estudio transversal utilizando el cuestionario PedsQL® para comparar la CVRS de 200 niños y adolescentes con obesidad al igual que a sus padres, con dos grupos de control, uno de la comunidad (n=400) y otro grupo de usuarios (n=200) del Hospital Infantil del Estado de Sonora. Las diferencias en la CVRS fueron eva...

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Pulpotomía con electrocauterio en pacientes con hemofilia A

OpenAIRE

Carbonell Rodriguez, Rosario

2017-01-01

Rehabilita la salud oral de un niños con Hemofilia A, empleando el Electrocauterio como una alternativa en la terapia pulpar, específicamente en la pulpotomía, para lo cual se realiza un minucioso examen clínico y anamnesis, que respaldados con exámenes complementarios, permite determinar el plan de tratamiento. La rehabilitación integral se llevó a cabo, previo consentimiento informado del padre, en sala de operaciones bajo anestesia general, debido a la condición sistémica del niño. Incl...

Intervención con menores con adicciones desde Proyecto Hombre

OpenAIRE

Pardo-Esteban, Belén; Escalza-González, Amaia

2015-01-01

En el citado seminario informaremos del programa de atención y prevención de las toxicomanías y otras adicciones, llevado a cabo por la Asociación Proyecto hombre. A continuación, trataremos la formación en adicciones y toxicomanía con los estudiantes, mediante charla, cuestionarios sobre conocimientos básicos en prevención de toxicomanía en menores y los diferentes programas de intervención posibles aducados a los perfiles de los usuarios. Terminaremos con un debate sobre los temas tra...

NARRATIVAS CONVERSACIONALES CON FAMILIAS Y DOCENTES DE NIÑOS Y NIÑAS CON DISCAPACIDAD: UN APORTE METODOLÓGICO

Directory of Open Access Journals (Sweden)

NORA ANETH PAVA-RIPOLL

2015-01-01

Full Text Available Se presentan algunas ventajas y limitaciones de las narrativas conversacionales como estrategia de investigación para el trabajo con familias y docentes de niños con discapacidad. La narración realizada por muchas voces, como sucede en una conversación, privilegia la relación y el discurso entre las personas como fuente de construcción consensuada de significados. La interacción triádica entre familia, inclusión educativa y salud orientó los encuentros conversacionales con padres, acudientes y profesores de estos niños en la ciudad de Manizales (Colombia. Los participantes conformaron equipos conversacionales y equipos reflexivos y conversaron sobre preguntas propuestas, previamente elaboradas. Con este ejercicio investigativo se resalta el valor interaccional de esta estrategia desde tres ámbitos: entre los par ticipantes, con el investigador y con el contexto situacional. Se argumenta que la reflexividad está más relacionada con el hacer que con el ser y el sentir de las personas, las cuales no necesariamente se transforman en la narrativa...

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%

Pilas de combustible de una sola cámara, basadas en electrolitos de ceria dopada con gadolinia y operadas con metano y propano

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Piñol, S.

2010-02-01

Full Text Available The main advantages of single-chamber solid oxide fuel cells (SOFCs respect to dual-chamber SOFCs, are to simplify the device design and to operate in mixtures of hydrocarbon (methane, propane… and air, with no separation between fuel and oxidant. However, this design requires the use of selective electrodes for the fuel oxidation and the oxidant reduction. In this work, electrolyte-supported SOFCs were fabricated using gadolinia doped ceria (GDC as the electrolyte, Ni + GDC as the anode and LSC(La_0.5Sr_0.5CoO_3-δ-GDC-Ag₂O as the cathode. The electrical properties of the cell were determined in mixtures of methane + air and propane + air. The influence of temperature, gas composition and total flow rate on the fuel cell performance was investigated. As a result, the power density was strongly increased with increasing temperature, total flow rate and hydrocarbon composition. Under optimized gas compositions and total flow conditions, power densities of 70 and 320 mW/cm² operating on propane at a temperature of 600ºC and methane (795ºC were obtained, respectively.

La principal ventaja de las pilas de combustible de óxido sólido (SOFCs de una sola cámara, frente a las bicamerales convencionales, es que permiten simplificar el diseño del dispositivo y operar con mezclas de hidrocarburos (metano, propano... y aire, sin necesidad de separar ambos gases, por medio del uso de electrodos selectivos a la oxidación del combustible y reducción del oxidante. En el presente trabajo, se han fabricado monopilas soportadas sobre electrolitos de ceria dopada con gadolinia (GDC, de 200 µm de espesor, usando Ni-GDC como ánodo y LSC(La_0.5Sr_0.5CoO_3-δ-GDC-Ag₂O como cátodo. Las propiedades eléctricas de la celda se determinaron en un reactor de una sola cámara, usando mezclas de metano + aire y propano + aire. Se investigó la influencia de la

MODELADO CON REDES DE PETRI E IMPLEMENTACIÓN CON GRAFCET DE UN SISTEMA DE MANUFACTURA FLEXIBLE CON PROCESOS CONCURRENTES Y RECURSOS COMPARTIDOS

Directory of Open Access Journals (Sweden)

Johanna Stella Castellanos Arias

2010-01-01

Full Text Available En este trabajo, se presenta el modelado de un Sistema de Manufactura Flexible (SMF, con procesos concurrentes y recursos compartidos mediante Sistemas a Eventos Discretos (SED, específicamente Redes de Petri (RdP, y GRAFCET. El SMF se plantea como un modelo hipotético que se modela con una RdP con el objeto de identificar su dinámica y hallar la secuencia óptima de funcionamiento del sistema. Se desarrolló un modelo matemático que permite estimar el vector de tiempo acumulado de un proceso modelado mediante una RdP, el cual constituye la base para hallar la mejor secuencia posible del sistema modelado. Por último, se realizó una implementación en el Laboratorio de Automatización de la Universidad Militar Nueva Granada, que simula el funcionamiento del SMF modelado por un montaje electrohidroneumático controlado con GRAFCET mediante PLC.

Relación del perfil salival con el grado de inmunosupresión en pacientes infectados con VIH con y sin tratamiento antirretroviral

OpenAIRE

Vera Cruz, Moisés; Cornejo Salaza, José; Chiyong, Teresa Evaristo; Arevalo Abanto, Jorge A.; Villanueva Vílchez, Hugo G.

2014-01-01

El objetivo del estudio fue evaluar el perfil salival de pacientes con infección por VIH con y sin Tratamiento Antirretroviral de Gran Actividad (TARGA), por medio de la evaluación de los síntomas de hipofunción salival (Xerostomía, Disgeusia, Disfagia y Susceptibilidad de desarrollar úlceras orales), la Tasa de Flujo salival no estimulado, el pH salival y la viscosidad salival; estableciendo su relación con el Grado de Inmunosupresión, evaluado por medio del Recuento de Linfocitos CD4/μl. Se...

Do existing research reactors teach us all about beam tube optimization?

International Nuclear Information System (INIS)

Roegler, Hans Joachim; Feltes, Wolfgang

1998-01-01

The contribution makes the attempt to analyse the data base available in the literature and in Siemens' own projects and to find out potential systematics from the existing research reactor with beam tubes, separated into reactors with different reflectors and distinguished for tangential and radial tubes and cold neutron sources, resp. Some generic calculations serve as gauging data. The contribution is not meant as critics on any design.The results might serve supporting designers and operators when evaluating the pros and cons of existing or planned design in terms of the optimum beam tubes. Existing lacks of systematics are evaluated in view of suitable explanations and constraints, which do not allow optimisation. Examples pf such constraints are the different material layers between fuel zone and reflector zone which have various reasons. The limited data in the literature plus the numerous lacks of precision of the representation of those data should be an incentive to improve the performed analysis by collecting more exact data and re-doing the evaluation before answering the title-question really. (author)

Do existing research reactors teach us all about beam tube optimisation?

International Nuclear Information System (INIS)

Roegler, Hans-Joachim; Feltes, Wolfgang

1998-01-01

The contribution makes the attempt to analyse the data base available in the literature and in Siemens' own projects and to find out potential systematics from the existing research reactors with beam tubes, separated into reactors with different reflectors and distinguished for tangential and radial tubes and cold neutron sources, respectively some generic calculations serve as gauging data. The contribution is not meant as critics on any design. The results might serve supporting designers and operators when evaluating the pros and cons of existing or planned design in terms of the optimum beam tubes. Existing lacks of systematics are evaluated in view of suitable explanations and constraints, which do not allow optimisation. Examples of such constraints are the different material layers between fuel zone and reflector zone which have various reasons. The limited data in the literature plus the numerous lacks of precision of the representation of those data should be an incentive to improve the performed analysis by collecting more exact data and re-doing the evaluation before answering the title question really

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

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

SATISFACCIÓN CON EL TRATAMIENTO EN PACIENTES DE ATENCIÓN PRIMARIA CON ARTROSIS

Directory of Open Access Journals (Sweden)

Carmen Llanos Val Jiménez

2017-01-01

Full Text Available Fundamentos: Al igual que en otros problemas crónicos, en el abor - daje del paciente con artrosis se tiende a un modelo de toma de decisiones compartidas, en el que el punto de vista del paciente debe ser considerado en la elección del tratamiento. El objetivo de este trabajo fue evaluar la sa - tisfacción con el tratamiento en pacientes diagnosticados de osteoartrosis y comprobar si existen diferencias entre diferentes opciones farmacológicas, así como determinar si existe asociación con el estado funcional y las carac - terísticas clínicas y sociodemográficas. Métodos: Se realizó un estudio observacional transversal en una muestra de 487 pacientes seleccionados mediante muestreo consecutivo. La variable principal fue el nivel de satisfacción con el tratamiento (cuestionario ARTS. El estado funcional de los pacientes fue evaluado mediante la escala WO - MAC. Otras variables fueron: características del tratamiento, adherencia te - rapéutica, eventos adversos, y variables clínicas y sociodemográficas. Resultados: En el cuestionario ARTS los pacientes, en un rango entre 28 y 87, obtuvieron una puntuación media de 65,3 (DE: 9,9. La puntuación no fue significativamente diferente en consumidores de 1, 2 o más fármacos. Entre quienes consumían un solo fármaco, no hubo diferencias entre los di - ferentes tipos de fármacos. En el estado funcional se obtuvo una puntuación media de 30,2 puntos (DE: 20,8 y se observó una débil correlación negativa con el nivel de satisfacción (r= - 0,252; p<0,001. Mediante regresión lineal múltiple, se observó mayor puntuación en la escala ARTS (p<0,05 en pa - cientes con menor puntuación en la escala WOMAC, mayor edad y ausencia de eventos adversos. Conclusión : En pacientes con osteoartritis se observa un nivel mode - rado de satisfacción con el tratamiento farmacológico, condicionado por su situación funcional, sus características sociodemográficas y por la presencia de eventos

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Hidroxilación de fenol con catalizadores ZSM-5 modificado con cobre

Directory of Open Access Journals (Sweden)

César Augusto Caro

2005-01-01

Full Text Available Se sintetizaron varios catalizadores Cu-ZSM-5 con diferentes relaciones Si/Al y Si/Cu. Usando metilamina o hidróxido de sodio como agente mineralizante. Los catalizadores se caracterizaron por DRX, IR, análisis BET, UV-VIS y análisis elemental. El desempeño catalítico de los catalizadores sintetizados se evaluó en la hidroxilación de fenol con H2O2. Se encontró que la relación catecol (CAT/(hidroquinona (HQ + p-benzoquinona (PBQ aumentó con el contenido de agua, presentó un máximo cuando la relación en peso de agua/fenol fue de 53/1 y se favoreció a altas temperaturas, con el aumento del contenido de aluminio o con la disminución en el contenido de catalizador. El contenido de cobre no fue un factor determinante para la conversión ni para la selectividad en la hidroxilación de fenol. La producción de la p-benzoquinona (PBQ, producto de oxidación de la hidroquinona, fue mínimo cuando se usaron las siguientes condiciones: 80 ºC, fenol: 1 mmol, relación molar fenol/H2O2 de 3/1, catalizador: 20 mg, agua: 5 g, tiempo de reacción: 4 h.

ESTUDIO DE LA DIGESTION HUMANA CON ALUMNOS CON DIFICULTADES DE APRENDIZAJE

Directory of Open Access Journals (Sweden)

Mª Begoña Carretero Gómez

2009-01-01

Full Text Available En esta experiencia se han seleccionado una serie de actividades relacionadas con la digestión con el fin de lograr una reestructuración de ideas y cambios conceptuales, y así con la conseguir que el alumnado aprenda a construir sus propios conocimientos acerca de cómo y donde se dan los procesos de digestión necesarios para la nutrición del organismo. El trabajo se ha llevado a cabo con alumnos de 3º de Educación Secundaria Obligatoria, pertenecientesal programa de diversificación curricular. Hemos querido hacer la ciencia más cercana para que comprendan que algunos procesos tienen su explicación científica y que la ciencia forma parte de nuestro día a día, fomentando el interés del estudiante así como su implicación en el proceso de enseñanza-aprendizaje. Por ello hemos trabajado a base de pequeñas investigaciones, donde es el alumno el que marca su ritmo de trabajo, investiga y hace sus aportaciones. Han aprendido a trabajar de una forma meticulosa donde el rigor esfundamental, contagiándose del afán investigador de manera que al realizar las experiencias se han aproximado al trabajo científico a la vez que aclaran y reestructuran conocimientos.

Estudio cinético de la reacción del CO con el O2 sobre un catalizador de Pd en alto vacío

Directory of Open Access Journals (Sweden)

Luis M. Carballo Suárez

1988-01-01

Full Text Available La oxidación del monóxido de carbono con el oxígeno sobre una banda de Paladio policristalino bajo condiciones de estado estacionario se investigó con la ayuda de un espectrómetro de masas y con la utilización de monóxido de carbono marcado isotópicamente 13CO. A temperaturas entre 500 y 750 K se varía la proporción de las presiones parciales de los reactantes entre los límites 0,01 hasta 100, y la presión total en el reactor está siempre por debajo de 10-3 mbar. Los resultados experimentales se describen usando una secuencia de pasos elementales que incluye la adsorción y desorción molecular de los reactantes, y la reacción de las dos especies adsorbidas para formar CO2, siendo este último el paso controlante de la reacción. El modelo considera dos clases diferentes de sitios activos. Los cálculos basados sobre valores de los parámetros tomados de la literatura muestran un ajuste aproximado con los datos experimentales.

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

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Esporotricosis cutánea diseminada con compromiso articular en una mujer con diabetes tipo 2

Directory of Open Access Journals (Sweden)

Sendy Solorzano

Full Text Available La esporotricosis diseminada es una presentación infrecuente de esta micosis. Aunque ha sido descrita en sujetos inmunocompetentes, con frecuencia existe compromiso de la inmunidad mediada por células T. Reportamos el caso de una mujer con diabetes tipo 2 no controlada que desarrolló esporotricosis cutánea diseminada. El curso de la enfermedad presentó varias complicaciones que incluyeron hepatitis secundaria a itraconazol y artritis de la rodilla con cultivo positivo para Sporothrix schenckii durante el tratamiento con solución saturada de ioduro de potasio. La discusión del caso incluye aspectos de la patogénesis de la esporotricosis diseminada y el manejo de la infección y sus complicaciones.

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)

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

BWR type reactor

International Nuclear Information System (INIS)

Watanabe, Shoichi

1983-01-01

Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

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.)

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

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

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.

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

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.)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

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.)

A multi-purpose reactor

International Nuclear Information System (INIS)

Changwen Ma

2000-01-01

An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

Efectos del tratamiento con litio sobre las funciones neurosicológicas de pacientes con trastorno bipolar de tipo I

Directory of Open Access Journals (Sweden)

Anabel Martínez Aran

2008-11-01

Full Text Available OBJETIVO: determinar si el tratamiento continuo con litio influye en el desempeño cognitivo de un grupo de pacientes eutímicos con trastorno afectivo bipolar I. MÉTODO: se evaluó el desempeño cognitivo usando pruebas neurosicológicas de atención, memoria y función ejecutiva en 20 pacientes eutímicos con trastorno afectivo bipolar I que no tomaban medicación, 20 pacientes eutímicos con trastorno afectivo bipolar I quienes continuaban regularmente el tratamiento con carbonato de litio, y un tercer grupo de 20 pacientes control. RESULTADOS: no se encontraron diferencias estadísticamente significativas entre los dos grupos de pacientes con trastorno afectivo bipolar. Sin embargo, al compararlos con el grupo control, su desempeño en términos de memoria verbal, viso-verbal y asociación semántica medida fue menor. DISCUSIÓN: los pacientes con trastorno afectivo bipolar I muestran deficiencias en términos de memoria verbal y viso-verbal que no pueden ser atribuidos al tratamiento regular con carbonato de litio. Por el contrario, estas deficiencias podrían ser consideradas como características de rasgo de este trastorno.

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

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

Conceptual design of a nucleo electric simulator with PBMR reactor based in Reduced order models; Diseno conceptual de un simulador de nucleo electrica con reactor PBMR basado en modelos de orden reducido

Energy Technology Data Exchange (ETDEWEB)

Valle H, J.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

2005-07-01

This project has as purpose to know to depth the operation of a PBMR nucleo electric type (Pebble Bed Modular Reactor), which has a reactor of moderate graphite spheres and fuel of uranium dioxide cooled with Helium and Brayton thermodynamic cycle. The simulator seeks to describe the dynamics of the one process of energy generation in the nuclear fuel, the process of transport toward the coolant one and the conversion to mechanical energy in the turbo-generators as well as in the heat exchangers indispensable for the process. The dynamics of reload of the fuel elements it is not modeled in detail but their effects are represented in the parameters of the pattern. They are modeled also the turbo-compressors of the primary circuit of the work fluid. The control of the power of the nuclear reactor is modeled by means of reactivity functions specified in the simulation platform. The proposed mathematical models will be settled in the platform of simulation of Simulink-Mat Lab. The proposed control panels for this simulator can be designed and to implement using the box of tools of Simulink that facilitates this process. The work presents the mathematical models more important used for their future implementation in Simulink. (Author)

Acne polimorfo: tratamiento con Implacen

Directory of Open Access Journals (Sweden)

Rubén Pérez Armas

1995-06-01

Full Text Available Se realiza el estudio de 40 pacientes con acné polimorfo, los que fueron atendidos en la Consulta de Dermatología del Hospital Provincial Clinicoquirúrgico Docente "Celia Sánchez Manduley", en el período comprendido de enero de 1988 a diciembre de 1989. Se revisa la literatura médica sobre los diversos métodos y medicamentos utilizados en la terapéutica de esta dermatosis. Se describe el esquema de tratamiento empleado con implacén en 30 pacientes; los 10 restantes se trataron con placebo; se compara dicho esquema con los tradicionales y se observan mejores resultados con nuestro estudio. Se destaca la ausencia de recaídas, así como el resultado del tratamiento de acuerdo con el sexo.A study was performed in 40 patients presenting with polymorphic acne who were attended in the Dermatology Department of "Celia Sánchez Manduley" Clinicosurgical and Teaching Hospital from January, 1988 to December, 1989. A review of the literature was made seeking for the different methods and drugs used for the treatment of this dermatosis. The treatment schedule with the use of implacen in 30 patients is described. Such therapeutic schedule was compared with traditional ones and better results were observed with the use of implacen. The fact that there were no relapses is highlighted, as well as the result of treatment according to sex.

Reducción de color con cloruro de magnesio en soluciones con colorantes comerciales

Directory of Open Access Journals (Sweden)

Mercedes Lucero Chávez

2017-01-01

Full Text Available Se experimentó con diferentes concentraciones de cloruro de magnesio (MgCl2 como coagulante para reducir el color en soluciones con colorantes comerciales: negro, azul y café. Los parámetros analizados fueron pH, color, demanda química de oxígeno ( dqo y sólidos suspendidos totales (sst de acuerdo con las Normas Mexicanas. Como prueba complementaria se agregó carbón activado a las muestras para reducir el color residual después de la coagulación-floculación. La reducción de los valores de color y dqo fue mayor cuando se adicionó más cantidad de MgCl2 (1.8 g/L. Los porcentajes de reducción fueron ≥90% para el color y >50% para la dqo. Los sst aumentaron con la adición de MgCl2. El carbón activado eliminó el color residual.

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

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

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

Directory of Open Access Journals (Sweden)

Ramiro Escalera Vásquez

2009-01-01

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

Aprende Ajedrez con Rey - Parte 2

OpenAIRE

ESTÉVEZ MONTERO, RAÚL; Lloret Mauri, Jaime

2016-01-01

Es una pieza audiovisual creada con el objeto de atraer la atención de los niños de muy corta edad con el ajedrez y familiarizarlos con todas sus piezas y movimientos. Es una animación dirigida a un público infantil presentada por dibujos animados en 2D, en la que se ha intentado respetar en todo momento el argot de la comunidad ajedredística. En este video se presenta la segunda parte. Estévez Montero, R.; Lloret Mauri, J. (2016). Aprende Ajedrez con Rey - Parte 2. http://hdl.handle.net/1...

Aprende Ajedrez con Rey - Parte 1

OpenAIRE

ESTÉVEZ MONTERO, RAÚL; Lloret Mauri, Jaime

2016-01-01

Es una pieza audiovisual creada con el objeto de atraer la atención de los niños de muy corta edad con el ajedrez y familiarizarlos con todas sus piezas y movimientos. Es una animación dirigida a un público infantil presentada por dibujos animados en 2D, en la que se ha intentado respetar en todo momento el argot de la comunidad ajedredística. En este video se presenta la primera parte. Estévez Montero, R.; Lloret Mauri, J. (2016). Aprende Ajedrez con Rey - Parte 1. http://hdl.handle.net/1...

Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

International Nuclear Information System (INIS)

Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

2013-01-01

The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

Design guide for Category III reactors: pool type reactors

International Nuclear Information System (INIS)

Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

1978-11-01

The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

Desalination of seawater with nuclear power reactors in cogeneration; Desalacion de agua de mar con reactores nucleares de potencia en cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Flores E, R M

2004-07-01

The growing demand for energy and hydraulic resources for satisfy the domestic, industrial, agricultural activities, etc. has wakened up the interest to carry out concerning investigations to study the diverse technologies guided to increase the available hydraulic resources, as well as to the search of alternatives of electric power generation, economic and socially profitable. In this sense the possible use of the nuclear energy is examined in cogeneration to obtain electricity and drinkable water for desalination of seawater. The technologies are analysed involved in the nuclear cogeneration (desalination technology, nuclear and desalination-nuclear joining) available in the world. At the same time it is exemplified the coupling of a nuclear reactor and a process of hybrid desalination that today in day the adult offers and economic advantages. Finally, the nuclear desalination is presented as a technical and economically viable solution in regions where necessities of drinkable water are had for the urban, agricultural consumption and industrial in great scale and that for local situations it is possible to satisfy it desalinating seawater. (Author)

The computerized reactor period measurement system for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1996-01-01

The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

Computer measurement system of reactor period for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1997-01-01

The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

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

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

1989-01-01

In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

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)

Videojuego con Realidad Virtual

OpenAIRE

González Mora, César

2017-01-01

El objetivo del proyecto es el desarrollo de un videojuego deportivo que utilice realidad mixta. El videojuego se podrá utilizar con dispositivos de tipo cardboard, y utilizará realidad aumentada para la interacción del jugador con el videojuego. En el desarrollo se utilizará el motor Unity para conseguir una aplicación multiplataforma, y la librería Vuforia para implementar realidad mixta.

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

Factores psicosociales relacionados con la rehospitalización de pacientes con psicosis orgánica

Directory of Open Access Journals (Sweden)

Cortés-Padilla María Teresa

2001-01-01

Full Text Available Objetivo. Conocer los factores psicosociales relacionados con la frecuente rehospitalización de pacientes con psicosis orgánica. Material y métodos. Estudio de tipo observacional descriptivo, de corte transversal. La muestra estuvo formada por 33 pacientes del Hospital Psiquiátrico Guillermo Dávila, con Unidad de Medicina Familiar Número 10 del Instituto Mexicano del Seguro Social de la Ciudad de México, con psicosis orgánica y quienes durante los años de 1993-1994 presentaron más de dos rehospitalizaciones. Instrumentos: Entrevista grabada, o muestra de conversación de cinco minutos, para la emoción expresada (EE, cuestionarios mixtos para la Concepción de Enfermedad y la Relación médico-paciente. La validez y confiabilidad de los datos se establecieron con las pruebas de Kappa y Alpha de Cronbach y se efectuó análisis estadístico descriptivo e inferencial. Resultados. El 60.9% de los familiares presentan alto nivel de EE, es decir, expresan crítica, hostilidad o sobreinvolucramiento emocional; el 64.3% de los sujetos de la muestra conviven más de 35 horas a la semana con familiares que presentan Alta EE. Conclusiones. Altos niveles de EE estuvieron asociados con los frecuentes reingresos a hospitalización. Existe mayor conocimiento de las características de la enfermedad por parte del familiar que del paciente. La relación médico-paciente es satisfactoria pero no forma conciencia de enfermedad ni propicia apego al tratamiento terapeútico. El texto completo en inglés de este artículo está disponible en: http://www.insp.mx/salud/index.html

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

Analysis of dynamic stability and safety of reactor system by reactor simulator

International Nuclear Information System (INIS)

Raisic, N.

1963-11-01

In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

Nuclear reactor physics

CERN Document Server

Stacey, Weston M

2010-01-01

Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

TRATAMIENTO DE AGUAS RESIDUALES MEDIANTE REACTORES ANAERÓBICOS DE PLACAS VERTICALES PARALELAS EN ACRÍLICO TRATAMENTO DE ÁGUAS RESIDUÁRIAS POR REATORES ANAERÓBIOS DE PLACAS VERTICAIS PARALELAS EM ACRÍLICO WASTEWATER TREATMENT BY ANAEROBIC REACTORS OF VERTICAL PARALLEL PLATES IN ACRYLIC

Directory of Open Access Journals (Sweden)

Guillermo Chaux F

2011-12-01

Full Text Available Algunos filtros anaeróbicos con lecho de piedra construidos en el departamento del Cauca (Colombia, están presentando problemas de colmatación. Si se reemplaza la piedra por placas verticales paralelas, se elimina el problema de obstrucción. Este documento presenta el desarrollo y resultados de una investigación que evaluó a escala de laboratorio el potencial de los reactores anaeróbicos de placas verticales paralelas en acrílico para remover contaminantes (materia orgánica y sólidos suspendidos. El reactor anaeróbico de placas paralelas en acrílico se desempeñó como tratamiento secundario; se alimentó con agua residual efluente de un Tanque Imhoff con concentraciones medias de 156 ± 14 mg/L de DB05, 438 ± 32 mg/L de DQO y 98 ±22 mg/L de sólidos suspendidos totales. Las remociones de DQO y DB05 en el reactor sobrepasan el 50% y la remoción de sólidos suspendidos sobrepasó el 60% para tiempos de detención de 24 horas. La facilidad en la operación del reactor lo hace viable como tratamiento biológico anaeróbico de aguas residuales previamente decantadasAlguns filtros anaeróbios com recheio de pedras construída no departamento de Cauca (Colombia estão apresentando problemas de obstrução. Se a pedra é substituída por placas verticais paralelas, evita o problema da obstrução. Este artigo apresenta o desenvolvimento e os resultados e no estudo realizado em escala de laboratório que avaliaram o potencial de reatores anaeróbios de placas verticais paralelas em acrílico para remover os contaminantes (sólidos suspensos e matéria orgânica. 0 reator anaeróbio de placas paralelas de acrílico serviu como tratamento secundário; foi alimentado com água residuária do efluente de um tanque Imhoff com concentrações médias de 156 ± 14 mg/L DB05, 438 ± 32 mg/L de DQO e 98 ±22 mg/L de sólidos suspensos totais. A remoção de DQO e DB05 no reator são mais de 50% ea remoção de sólidos em suspensão superior a 60

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

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.

Argentinean integrated small reactor design and scale economy analysis of integrated reactor

International Nuclear Information System (INIS)

Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

2000-01-01

This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

Análisis probabilista con caminata al azar del número de personas viviendo con VIH mundialmente

Directory of Open Access Journals (Sweden)

Javier Oswaldo Rodríguez-Velásquez

2018-01-01

Full Text Available Introducción: El estudio de la dinámica anual de la epidemia de VIH a partir de la caminata al azar ha mostrado ser de utilidad para llevar este fenómeno altamente variable a un comportamiento predecible Objetivo: Predecir el comportamiento de la dinámica del número de personas viviendo con VIH mediante la caminata al azar probabilista. Metodología: Se analizaron los valores anuales de personas que a nivel global viven con VIH desde el año 1990 hasta el 2009, a partir de los espacios de probabilidad generados con la caminata al azar probabilista, para posteriormente desarrollar la predicción del valor anual de personas viviendo con VIH para los años 2010, 2011 y 2012. Resultados: Se predijeron los valores anuales de personas viviendo con VIH con un porcentaje de acierto de 98,95% para el año 2010, de 98,82% para el año 2011 y de 98,99% para la predicción realizada para el año 2012. Conclusiones: Se establecieron órdenes matemáticos a partir de la caminata al azar probabilista, estableciendo predicciones prácticamente deterministas del número de personas viviendo con VIH, lo que podría ser útil para las decisiones en Salud Pública y para la evaluación de intervenciones.

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.)

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, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

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.)

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Fuerza manual de adultos con discapacidad intelectual

Directory of Open Access Journals (Sweden)

Ruth Cabeza Ruiz

2017-09-01

Full Text Available Objetivo. Presentar una descripción de la fuerza de prensión manual de hombres y mujeres con discapacidad intelectual (DI y comparar los resultados con valores de referencia de otras personas con y sin discapacidad intelectual. Método. El presente trabajo es un estudio transversal observacional, financiado por la Fundación SAMU, en el que se evaluaron a 122 personas con DI (86 hombres y 36 mujeres durante el desarrollo de unas jornadas de carácter recreativo en las que participaron varias asociaciones de atención a este colectivo. La batería de test utilizada fue el Alpaha-Fit Test Battery for Adults. Resultados. Se presentan los resultados relacionados con las variables de fuerza del miembro superior (Hand Grip Strength por grupos de edad (20-24, 25- 29, 30-34, 35-39, 40-44, 45-49, 50-54, 55-59 años. Los datos muestran valores que oscilan desde los 31 kg en los hombres más jóvenes con DI hasta los 13.3 kg del grupo más maduro de mujeres. Estos hallazgos son similares a los valores de referencia de población con DI española. Sin embargo, son muy inferiores a los obtenidos por la población sin discapacidad de la misma edad. Conclusión. Los resultados evidencian el menor rendimiento de las personas con DI en pruebas de fuerza de prensión manual por lo que se hace evidente la necesidad de llevar a cabo programas de ejercicio físico o deporte con las personas con DI.

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

Sistemas integrados con Arduino

OpenAIRE

EL YAKOUTI, MOHAMMED

2017-01-01

Design of a robot prototype remotely controllable from Bluetooth using Arduino. Control and testing of sensors and events interacting with Arduino and Bluetooth. Diseño de un prototipo de robot controlable remotamente con Bluetooth utilizando Arduino. Control y verificación de los sensores y eventos que interactúan mediante el Arduino y el Bluetooth. El Yakouti, M. (2017). Sistemas integrados con Arduino. http://hdl.handle.net/10251/89274. TFGM

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

To the analysis of reactor noise in boiling water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1972-01-01

The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

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

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

fertilizada con diferentes abonos

Directory of Open Access Journals (Sweden)

Jorge Alberto Elizondo-Salazar

2007-01-01

Full Text Available Producción y calidad de la biomasa de morera (Morus alba fertilizada con diferentes abonos. Se llevó a cabo un experimento en la Estación Experimental “Alfredo Volio Mata” de la Universidad de Costa Rica con el fi n de evaluar la aplicación de 150 kg de N/ha/año proveniente de dos abonos orgánicos: lombriabono y compostaje; y de un fertilizante químico, sobre la producción y calidad de la biomasa de morera. El periodo experimental comprendió un ciclo de 12 meses, iniciando en julio del 2003 y fi nalizando en julio del 2004. Se utilizó una plantación de morera de 12 años de establecida con una densidad de siembra de 27.777 plantas/ ha. Se empleó un diseño de bloques completos al azar con cuatro tratamientos: dos abonos orgánicos, nitrato de amonio (33,5% N y un control. Las plantas se podaron a 0,6 m sobre el nivel del suelo al inicio del ensayo. Durante el periodo experimental, las plantas fueron podadas consecutivamente cada 90 días. Las hojas y los tallos fueron separados y analizados para determinar el contenido de materia seca y proteína cruda. La producción de materia seca fue 23% superior y el contenido de proteína cruda fue signifi cativamente mayor con el nitrógeno químico, mientras que el contenido de materia seca fue menor. No se encontraron diferencias signifi cativas entre el tratamiento control y los tratamientos orgánicos.

Estudio observacional prospectivo con insulina detemir en pacientes con diabetes mellitus tipo 2 mal controlados que inician por primer vez tratamiento con insulina (estudio SOLVE

Directory of Open Access Journals (Sweden)

Domingo Orozco-Beltrán

2016-02-01

Conclusiones: En esta cohorte de pacientes con diabetes mellitus tipo 2 recientemente insulinizados, la insulina detemir (una vez al día mejoró el control glucémico, con baja incidencia de hipoglucemia y una reducción significativa del peso.

Eficacia del tratamiento tópico con Vimang® en pacientes con estomatitis subprótesis

Directory of Open Access Journals (Sweden)

Laritza Coureaux Rojas

2013-10-01

Full Text Available Se efectuó un ensayo clínico terapéutico controlado de fase III en 120 pacientes con estomatitis subprótesis, atendidos en la Clínica Estomatológica Provincial Docente "Mártires del Moncada" de Santiago de Cuba, desde junio de 2010 hasta mayo de 2011, con vistas a evaluar la eficacia del tratamiento tópico con té de Vimang® en dichos pacientes. Se conformaron 2 grupos de 60 integrantes cada uno (de estudio y de control. Los primeros fueron tratados con enjuagatorios fríos de dicho té 4 veces al día y los segundos con el mismo tratamiento más vitaminas. Se utilizaron el porcentaje para variables cualitativas, así como la media aritmética y la desviación estándar para variables cuantitativas. Todos los integrantes del grupo de estudio se curaron y disminuyó el gasto de material, pero el costo-efectividad fue mayor por ser menor el tiempo de curación. El té de Vimang® resultó eficaz para eliminar las manifestaciones clínicas de esa lesión y no hubo reacciones adversas, de manera que se logró la curación de los afectados con ahorro de tiempo y recursos.

Generation of nuclear constants of the TRIGA reactor with the Leopard code; Generacion de constantes nucleares del reactor TRIGA con el codigo Leopard

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F; Perusquia del C, R [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1983-09-15

The reactor core was divided in 12 regions, this was made in function of the composition and temperature and its are: 1) central thimble, 2) B ring, 3) C ring, 4) D ring, 5) E ring, 6) F ring, 7) G ring, 8) superior caps of fuel elements (E.C. s) standard, 9) inferior caps of E.C.'s standard, 10) superior and inferior reflector of the core, 11) lateral reflector and 12) superior and inferior caps of the E.C.'s graphite. Likewise the constants of the followers' of fuel cell, of the empty follower and of the conduits of the gamma camera were obtained. For the obtaining of the enter data of the LEOPARD the dimensions and the composition of the different regions are required, this is consigned in the IT/E21-83 report. (Author)

Virtual nuclear reactor for education of nuclear reactor physics

International Nuclear Information System (INIS)

Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

2008-01-01

As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

Trastornos temporomandibulares en pacientes con maloclusiones

Directory of Open Access Journals (Sweden)

Luis Soto Cantero

Full Text Available Introducción: existe una prevalencia de trastornos oclusales en gran medida relacionados con la presencia de disfunción temporomandibular. Objetivo: determinar el grado de disfunción temporomandibular según el Índice de Maglione en correspondencia con la prevalencia de maloclusiones en el área de salud del policlínico "Turcios Lima". Métodos: se realizó un estudio observacional descriptivo, de corte transversal, en la consulta de ortodoncia, de junio del 2008 a junio del 2009. De un universo constituido por los 280 pacientes que acudieron al servicio de ortodoncia en el período señalado con presencia de maloclusiones, se seleccionó una muestra de 84 pacientes por muestreo aleatorio simple, teniendo en cuenta los criterios de inclusión y exclusión realizados por criterios de expertos. Resultados: presentaban disfunción temporomandibular 74 pacientes y dentro de ellos 44 (52,4 % tenían disfunción grado II (Moderada. El 97 % de los pacientes con relación molar de clase II, presentaban disfunción temporomandibular. El 42,9 % de los pacientes con una maloclusión, presentaban disfunción grado I. El 60 % de los pacientes con dos maloclusiones presentaban disfunción grado II y el 66,7 % de los pacientes con tres maloclusiones, presentaban disfunción grado III. La mayor cantidad de pacientes tuvieron disfunción grado II (Moderada. Conclusiones: el mayor por ciento de los pacientes con disfunciones presentó una clase II molar y a medida que aumentó el número de maloclusiones aumentó también la severidad de la disfunción.

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.)

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

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

Energy Technology Data Exchange (ETDEWEB)

Kliem, S.

2017-09-15

The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

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)

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

RadCon: A Radiological Consequences Model

International Nuclear Information System (INIS)

Crawford, J.; Domel, R.U.

2000-05-01

RadCon estimates the dose received by user selected groups in the population from an accidental release of radionuclides to the environment. The exposure pathways considered are external exposure from the cloud and ground and internal exposure from inhalation and ingestion of contaminated food. Atmospheric dispersion modelling is carried out externally to RadCon.Given a two dimensional time varying air and ground concentration of radioactive elements, RadCon allows the user to: view the air and ground concentration over the affected area, select optional parameters and calculate the dose to people,display the results to the user, and change the parameter values. RadCon offers two user interfaces: 1) the standard graphical user interface which is started using Java DoseApp at the command line, or by setting up a shortcut to this command (particularly when RadCon is installed on a PC) and 2) the text based interface used to generate information for the model inter-comparison exercise . This is initiated using Java BIOMASS at the command line, or an equivalent shortcut. The text based interface was developed for research purposes and is not generally available. Appendices A, B and C provide a summary of instructions on setting up RadCon. This will generally be carried out by the computer support personnel

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1984-07-01

It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

Ulysse, mentor reactor

International Nuclear Information System (INIS)

Bouquin, B.; Rio, I.; Safieh, J.

1997-01-01

On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

LICOR DE MORA DE CASTILLA (Rubus glaucus Benth CON DIFERENTES PORCENTAJES DE PULPA BLACKBERRY LIQUOR (Rubus glaucus Benth WITH DIFFERENT PULP PERCENTAGES

Directory of Open Access Journals (Sweden)

Álvaro Montoya Gómez

2005-12-01

Full Text Available La mora de Castilla (Rubus glaucus Benth es una fruta de interés comercial y altamente perecedera, que presenta deterioro debido a su fragilidad e inadecuado manejo poscosecha, cantidades apreciables de fruta son afectadas, básicamente en sus características físicas, siendo este producto apto para el procesamiento industrial. Veinte kg de mora fueron despulpados, homogenizados y caracterizada la pulpa, determinando; grados brix de 6,0; porcentaje de acidez de 2,91 % y densidad de 0,991 g/ml en promedio. Se formularon licores, con una participación de la pulpa del 25 %, 30 %, 35 %, 40 %, 45 %, 50 % y 55 % iniciando con 35 °Brix, para todos los casos en el mosto fermentable, ajustando las formulaciones con sacarosa; la levadura inoculada fue Saccharomyces cerevisiae Meyen ex E.C. Hansen en concentración del 0,2 %; la fermentación se llevó a cabo en reactores con capacidad de 2 litros, con desfogue de manguera de látex, el mosto fermentable fue dejado en los reactores durante 20 semanas, luego los licores fueron filtrados y caracterizados físico-químicamente, evaluando el porcentaje de alcohol por destilación, encontrando que el porcentaje de 40 % de participación de pulpa, fue el de mayor concentración con 8,36 % de volumen de alcohol en promedio. Los resultados estadísticos mostraron una diferencia significativa entre los porcentajes de participación de pulpa y su influencia sobre la producción de alcohol para los valores de 25 % y 30 % con respecto a los porcentajes 35 %, 40 %, 45 %, 50 % y 55 %. Un modelo matemático lineal apropiado para la producción de alcohol en función de la participación de pulpa fue obtenido. La prueba sensorial con 10 jueces, mostró que el licor preparado con 35 % de participación de pulpa, fue el de mayor aceptación.The blackberry (Rubus glaucus Benth is a fruit of commercial interest that is highly perishable, showing appreciable damage due to fragileness and inadequate pos-harvesting handling

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

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.)

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

Science.gov (United States)

Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

2016-01-01

Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

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.)

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)

Development of fiber-delivered laser peening system to prevent stress corrosion cracking of reactor components

International Nuclear Information System (INIS)

Sano, Y.; Kimura, M.; Yoda, M.; Mukai, N.; Sato, K.; Uehara, T.; Ito, T.; Shimamura, M.; Sudo, A.; Suezono, N.

2001-01-01

The authors have developed a system to deliver water-penetrable intense laser pulses of frequency-doubled Nd:YAG laser through optical fiber. The system is capable of improving a residual stress on water immersed metal material remotely, which is effective to prevent the initiation of stress corrosion cracking (SCC) of reactor components. Experimental results showed that a compressive residual stress with enough amplitude and depth was built in the surface layer of type 304 stainless steel (SUS304) by irradiating laser pulses through optical fiber with diameter of 1 mm. A prototype peening head with miniaturized dimensions of 88 mm x 46 mm x 25 mm was assembled to con-firm the accessibility to the heat affected zone (HAZ) along weld lines of a reactor core shroud. The accessibility was significantly improved owing to the flexible optical fiber and the miniaturized peening head. The fiber delivered system opens up the possibility of new applications of laser peening. (author)

Fast Spectrum Reactors

CERN Document Server

Todd, Donald; Tsvetkov, Pavel

2012-01-01

Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1985-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

1982-09-01

Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

International Nuclear Information System (INIS)

Shiba, T.; Fallot, M.

2015-01-01

To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

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.

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Reactor control device

International Nuclear Information System (INIS)

Fukami, Haruo; Morimoto, Yoshinori.

1981-01-01

Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Estudio comparativo del efecto del cepillado con una crema dental con propóleos rojos y de un gel con clorofila

Directory of Open Access Journals (Sweden)

Estela Gispert Abreu

1998-12-01

Full Text Available Se analiza comparativamente el efecto de un gel dental de clorofila y de una crema dental con propóleos rojos sobre varios parámetros relacionados con la caries dental, en escolares que se cepillaron durante 21 días con dichos productos. Se obtuvieron resultados favorables principalmente en la disminución del grado de infección por Streptococcus mutans y la elevación de la capacidad individual de remineralización; salvo en este último no hubo diferencias estadísticamente significativas.Authors made a comparative analysis about effect of chlorophyl dental gel and toothpaste containing red propolis, on some parameters related to dental caries in a group of students that brushed their teeth during 21 days using such products. Favourable results were obtained, mainly in decrease of infection level from Streptococcus mutans, and in increase of individual capacity of remineralization; except for this latter, there weren´t statistically significant differences.

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

Tratamiento conservador en pacientes con retinoblastoma bilateral

Directory of Open Access Journals (Sweden)

Juan C. Suárez

2008-11-01

Full Text Available OBJETIVO: comparar el tratamiento convencional del retinoblastoma bilateral, usado hasta hace algunos años, consistente en radioterapia o enucleación bilateral, con el tratamiento conservador actual que incluye termoterapia transpupilar (TTT o TTT/quimioterapia al menos en un ojo, en niños con diagnóstico de retinoblastoma bilateral. DISEÑO: estudio retrospectivo descriptivo. MUESTRA: 20 pacientes con diagnóstico de retinoblastoma bilateral que consultaron al Hospital Universitario San Vicente de Paúl, de Medellín, Colombia, entre 1997 y 2007. MÉTODO: se hizo enucleación del ojo con el tumor de mayor tamaño. En el otro ojo se hizo tratamiento con TTT, con el láser diodo (810 nm, spot amplio, solo o combinado con otras terapias. RESULTADOS: se dividió a los pacientes en dos grupos: 16 pacientes (32 ojos en el grupo 1 tratados conservadoramente y 4 pacientes (8 ojos en el grupo 2 con tratamiento convencional. El rango de edad fue de 1-72 meses en el grupo 1 y de 1-12 meses en el grupo 2. El tiempo de seguimiento fue de 7-67 meses para el grupo 1 y de 13-73 meses para el grupo 2. En el grupo 1 se hizo enucleación de 16 ojos (50%, radioterapia externa de uno (3,1%, quimioterapia más termoterapia de 5 (15,6% y quimioterapia más termoterapia más crioterapia de 10 (31,3%. En todos los pacientes se logró preservar al menos un ojo. En el grupo 2, se enuclearon 7 ojos (87,5% y se hizo radioterapia externa más enucleación en un paciente (12.5%. Además, todos los pacientes recibieron quimioterapia. CONCLUSIÓN: la terapia conservadora actual consistente en tratamiento local (termoterapia, crioterapia o braquiterapia y quimiorreducción permite preservar al menos un ojo y en algunos casos de los dos, muchas veces con buena agudeza visual, en niños con retinoblastoma bilateral; se evitan así la enucleación bilateral y la radioterapia externa usada en el tratamiento convencional con todos sus efectos secundarios. La enucleación contin

Variables socionutricionales de hogares mazahuas integrados por preescolares desnutridos con madres con obesidad y sin obesidad

OpenAIRE

Viridiana Vanessa Conzuelo-González

2009-01-01

El primer objetivo fue conocer cuántos menores de cinco años con diferentes grados de desnutrición tienen una madre con sobrepeso/obesidad/ en una comunidad indígena que vive en extrema pobreza y bajo condiciones de migración masculina internacional. El segundo fue comparar tres variables socionutricionales (ingreso familiar, educación de la madre y adecuación nutrimental de la dieta diaria) entre estos hogares y los hogares con desnutrición infantil y madres sin obesidad. Se realizó un estud...

Terapia asistida con perros en niños con autismo revisión sistemática

OpenAIRE

Isaza Narváez, Ana Milena; Muñoz Chilito, Doumer Adrian

2013-01-01

Introducción: El autismo es un trastorno temprano y masivo del desarrollo en el cual se afecta la interacción social, el desarrollo del lenguaje, se presentan rituales y estereotipias. Hay evidencia que la interacción con un perro está asociada con efectos positivos en la salud del niño, lo cual se extiende al terreno terapéutico donde la compañía del perro estimula habilidades comunicativas para conectarse con el mundo exterior y romper el aislamiento que lo hace refractario a formas convenc...

Variables socionutricionales de hogares mazahuas integrados por preescolares desnutridos con madres con obesidad y sin obesidad

OpenAIRE

Viridiana Vanessa Conzuelo-González; Ivonne Vizcarra-Bordi

2009-01-01

El primer objetivo fue conocer cuántos menores de cinco años con diferentes grados de desnutrición tienen una madre con sobrepeso/obesidad/ en una comunidad indígena que vive en extrema pobreza y bajo condiciones de migración masculina internacional. El segundo fue comparar tres variables socionutricionales (ingreso familiar, educación de la madre y adecuación nutrimental de la dieta diaria) entre estos hogares y los hogares con desnutrición infantil y madres sin obesidad. Se realizó un est...

Reactor safety device

International Nuclear Information System (INIS)

Okada, Yasumasa.

1987-01-01

Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1980-09-01

Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

Efectos del contacto piel con piel del recién nacido con su madre

Directory of Open Access Journals (Sweden)

Camila Lucchini Raies

Full Text Available Antecedentes: el contacto piel con piel ha sido una práctica ancestral en toda cultura, que debido a la institucionalización de la atención del parto, disminuyó en occidente. Se retomó en el marco de la atención humanizada del parto. Objetivo principal: determinar los efectos del contacto piel con piel del recién nacido con su madre. Metodología: revisión bibliográfica de 26 estudios. Límites de búsqueda: new borns, humans, clinical trials, de 10 años de publicación, en bases de datos: Pubmed, ProQuest, CINHAL, Lilacs. Resultados principales: Se observaron beneficios en las variables: duración de lactancia materna, comportamiento neurosensorial, regulación de temperatura, como alternativa al cuidado estándar, disminución en días hospitalización y niveles de cortisol, favorecedor de la interacción madre-hijo y desarrollo infantil. Conclusión principal: el contacto piel con piel brinda múltiples beneficios a recién nacidos. Se presenta como alternativa de cuidado segura, factible de implementar.

Reactor scram device for FBR type reactor

International Nuclear Information System (INIS)

Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

1991-01-01

In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Somatic mutation and recombination induced with reactor thermal neutrons in Drosophila melanogaster; Mutacion y recombinacion somaticas inducidas con neutrones termicos de reactor en Drosophila melanogaster

Energy Technology Data Exchange (ETDEWEB)

Zambrano A, F.; Guzman R, J.; Paredes G, L.; Delfin L, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1997-07-01

The SMART test of Drosophila melanogaster was used to quantify the effect over the somatic mutation and recombination induced by thermal and fast neutrons at the TRIGA Mark III reactor of the ININ at the power of 300 k W for times of 30, 60 and 120 minutes with total equivalent doses respectively of 20.8, 41.6 and 83.2 Sv. A linear relation between the radiation equivalent dose and the frequency of the genetic effects such as mutation and recombination was observed. The obtained results allow to conclude that SMART is a sensitive system to the induced damage by neutrons, so this can be used for studying its biological effects. (Author)

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.)

Compilation of reactor physics data of the year 1984, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-12-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Compilation of reactor physics data of the year 1983, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-06-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

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.)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1978-10-01

Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment; Simulador de la cinetica puntual de un reactor nuclear TRIGA Mark III con control difuso de potencia en un ambiente visual

Energy Technology Data Exchange (ETDEWEB)

Perez M, C

2004-07-01

The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

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

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

Snoj, L.; Kromar, M.; Zerovnik, G.

2012-01-01

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Larvicultura de Rhamdia quelen (Pisces, Pimelodidae con proteína vegetal y animal, suplementadas con plancton

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Germán Castañeda A.

2011-12-01

Full Text Available Objetivo. Determinar el efecto de dietas con diferentes proporciones de fuentes proteicas animales y vegetales suplementadas con plancton sobre el crecimiento y sobrevivencia en larvas de Rhamdia quelen. Materiales y métodos. Fueron formuladas dos dietas experimentales, dieta-1 con 70% de proteína de origen vegetal (torta de soya y dieta-2 con 70% de proteína proveniente de animales (corazón bovino yharina de pescado; se utilizaron cuatro protocolos de alimentación (cinco réplicas por tratamiento: dieta-1 + plancton filtrado en tamiz de 50-200 micras (T1, dieta-2 + plancton filtrado en tamiz de 50-200 micras (T2, dieta-1 (T3, dieta-2 (T4. Las larvas fueron manejadas a una densidad de 20 animales L-1 suministrando alimentohasta aparente saciedad diariamente a las 07:00, 11:00, 16:00 y 21:00 horas; la suplementación con plancton se realizó a las 11:00 y 21:00 horas. Resultados. Entre T1 y T2 no se observaron diferencias estadísticas (p>0.05 en peso final (21.89 ± 15.17mg vs 20.37 ± 10.37mg, longitud total (13.41±2.34mm vs 13.39±1.99 mm, factor de condición (K (0.80±0.13 vs 0.78±0.13 y sobrevivencia (46.6±2.68% vs36.0±7.41%; las diferencias entre T3 y T4 tampoco fueron significativas (3.35±1.40 mg vs 2.98±1.48 mg; 7.54±0.91mm vs 7.33±0.96mm; 0.75±0.13 vs 0.71±0.12; 33.6±9.07% vs 24.8±6.76%, respectivamente; hubo diferencias significativas (p<0.05 entre los grupos suplementados con plancton y los alimentados solo con ración. Conclusiones. La suplementación con plancton fue más efectiva que elofrecimiento solo de ración. La inclusión del 62.9% de torta de soya en la formulación (T1, aparentemente no afectó el crecimiento ni la sobrevivencia.

Ensayo clínico con intervención educativa en mujeres perimenopáusicas con un factor de riesgo cardiovascular

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Anxela Soto-Rodríguez

2017-01-01

Conclusiones: Una sencilla intervención educativa en mujeres perimenopáusicas con un factor de riesgo cardiovascular mejora aspectos relacionados con la calidad de vida y con hábitos saludables como la actividad física.

Soporte social con el uso de tic’s para cuidadores familiares de personas con enfermedad crónica

OpenAIRE

Carrillo, Gloria Mabel; Chaparro, Olivia Lorena; Barrera, Lucy; Sanchez, beatriz; Pinto, Raquel Natividad

2012-01-01

Objetivo Determinar la efectividad del soporte social dado con el uso de tecnologías de la información y la comunicación (TIC’S) a los cuidadores familiares de personas con enfermedad crónica.  Material y Métodos. Estudio descriptivo, exploratorio realizado en Bogotá en 2008 y 2009 con 144 cuidadores familiares de personas con enfermedad crónica  que constó de tres etapas para su desarrollo: 1. Planeación, diseño de la TIC’s,  aplicación de  prueba piloto y  selección de  población objeto;  ...

Reactor feedwater system

International Nuclear Information System (INIS)

Hikabe, Katsumi.

1978-01-01

Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

New reactor concepts

International Nuclear Information System (INIS)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Fission reactors and materials

International Nuclear Information System (INIS)

Frost, B.R.T.

1981-12-01

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J. Jr.

1981-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

Method of operating a reactor

International Nuclear Information System (INIS)

Oosumi, Katsumi; Yamamoto, Michiyoshi.

1980-01-01

Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

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.)

INFLUENCIA DE LA CARGA ORGÁNICA SOBRE LA EFICIENCIA DE REACTORES RBC DE TRES ETAPAS EN EL TRATAMIENTO DE UN EFLUENTE INDUSTRIAL SINTÉTICO

Directory of Open Access Journals (Sweden)

Elisabeth Behling de Calmón

2012-01-01

Full Text Available El presente artículo incluye el estudio de la influencia de la carga orgánica (CO sobre la eficiencia de reactores biológicos rotativos de contacto (RBC, aerobios de tres etapas, al tratar un efluente industrial sintético, con la finalidad de establecer la adecuación del efluente final con respecto a los límites de descarga establecidos en la normativa de Venezuela. Durante la experimentación, se evaluaron pH, alcalinidad total, oxígeno disuelto, demanda biológica de oxígeno (DBO, demanda química de oxígeno (DQO, sólidos suspendidos totales (SST, nitrógeno total Kjeldahl (NTK, NH 4 + , NO 2 - y NO 3 - , de acuerdo con los métodos estándares. La variación de la CO aplicada se obtuvo mediante modificación del tiempo de retención hidráulico (TRH, (24, 12 y 6 h, y se mantuvo constante la DQO de entrada (influente sintético de sacarosa+ urea. La mayor eficiencia de remoción de DQO se obtuvo para un CO global de 11,68 gDQO/m 2 .d (96,25%; TRH=12 h. Para los TRH 24 y 12 h, la eficiencia de remoción global de N-total fue de 66,92 y 62,95%, respectivamente. La mayor remoción de C y de N se obtuvo en la primera etapa de los reactores y se logró cumplir con el límite venezolano permisible de descarga para DQO (<350 mg/L. La posible ocurrencia del proceso de nitrificación repercutió sobre el aumento de las concentraciones de nitrógeno inorgánico en el efluente final.

Adolescente femenino con granulomatosis de Wegener fulminante

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S C Scheffler-Mendoza

2017-01-01

Full Text Available Adolescente femenino de 13 años de edad, originaria de Guerrero, grupo étnico mixteco, padres analfabetos y con poco entendimiento del español. Sin otros antecedentes de importancia para el padecimiento. Inició un mes previo con la aparición de una pápula hipercrómica en párpado superior derecho que en los 10 días previos al ingreso aumentó de volumen generando proptosis. Al interrogatorio se negó dolor, lagrimeo, fiebre, o pérdida de peso. Ingresó con fiebre (39.0°C, palidez generalizada, proptosis derecha con exposición de córnea y conjuntiva, eritema periorbitario, movimientos oculares y agudeza visual disminuidos (i.e. contaba dedos a 1 m, pupila con tendencia a midriasis, hiperreactividad a la luz, y fondo de ojo con papila hiperémica edematosa y tortuosidad peripapilar. El resto de exploración no mostró datos relevantes. Se hospitalizó para iniciar tratamiento antimicrobiano parenteral de amplio espectro (i.e. ceftriaxona y vancomicina y abordar proptosis unilateral.

Design of a homogeneous subcritical nuclear reactor based on thorium with a source of californium 252; Diseno de un reactor nuclear subcritico homogeneo a base de Torio con una fuente de Californio 252

Energy Technology Data Exchange (ETDEWEB)

Delgado H, C. E.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: ce_delgado89@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of)

2015-10-15

Full text: One of the energy alternatives to fossil fuels which do not produce greenhouse gases is the nuclear energy. One of the drawbacks of this alternative is the generation of radioactive wastes of long half-life and its relation to the generation of nuclear materials to produce weapons of mass destruction. An option to these drawbacks of nuclear energy is to use Thorium as part of the nuclear fuel which it becomes in U{sup 233} when capturing neutrons, that is a fissile material. In this paper Monte Carlo methods were used to design a homogeneous subcritical reactor based on thorium. As neutron reflector graphite was used. The reactor core is homogeneous and is formed of 70% light water as moderator, 12% of enriched uranium UO{sub 2}(NO{sub 3}){sub 4} and 18% of thorium Th(NO{sub 3}){sub 4} as fuel. To start the nuclear fission chain reaction an isotopic source of californium 252 was used with an intensity of 4.6 x 10{sup 7} s{sup -1}. In the design the value of the effective multiplication factor, whose value turned out k{sub eff} <1 was calculated. Also, the neutron spectra at different distances from the source and the total fluence were calculated, as well as the values of the ambient dose equivalent in the periphery of the reactor. (Author)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1993-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor engineering department annual report

International Nuclear Information System (INIS)

1990-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Síntomas depresivos en pacientes con Trastorno por déficit de atención con hiperactividad.

Directory of Open Access Journals (Sweden)

Diego Artemio Reátegui Villegas

2008-07-01

Full Text Available Objetivos: Determinar la frecuencia de síntomas depresivos en pacientes con Trastorno por déficit de atención con hiperactividad (TDAH, así como determinar su asociación con edad, sexo y comorbilidad psiquiátrica. Material y Método: Se realizó un estudio descriptivo de corte transversal. Se incluyeron pacientes con diagnóstico de TDAH del Instituto Nacional de Salud Mental Honorio Delgado Hideyo Noguchi, desde febrero del 2006 hasta completar el tamaño de la muestra. Se les aplicó a los pacientes el Children´s depression inventory (CDI. Se consideró un punto de corte de 15 para definir presencia de trastorno depresivo. Se analizaron las siguientes variables: edad, sexo, comorbilidad psiquiátrica y sintomatología depresiva. Resultados: Sesenta y cuatro (85,3% fueron varones, en 35/75 (46,7% se encontró CDI ³ 15. No se encontró asociación estadísticamente significativa entre síntomas depresivos y sexo o edad. El síntoma depresivo más frecuente fue el problema del rendimiento escolar (65,3%. Se encontraron 6 pacientes con otra comorbilidad. Conclusión: Existe una alta frecuencia de sintomatología depresiva en niños y adolescentes con diagnóstico de TDAH. (Rev Med Hered 2008;19:96-101 .

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

1977-09-01

Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1984-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1976-09-01

Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

Caracterización de pacientes con lesiones pigmentarias de piel con riesgo de transformación maligna

Directory of Open Access Journals (Sweden)

María Elena Izquierdo Izquierdo

2015-06-01

Full Text Available Fundamento: el melanoma es el más terrible y mortífero de los cánceres de piel. Se origina frecuentemente a partir de un nevo melanocítico con riesgo de transformación maligna. Estas lesiones pigmentadas se diagnostican cada vez más en edades tan tempranas como la adolescencia. Objetivo: caracterizar los pacientes en edades pediátricas con signos de lesiones pigmentarias de la piel con riesgo de transformación maligna. Métodos: se realizó un estudio prospectivo de corte transversal en un grupo de 50 pacientes en edades entre 5 y 19 años, que asistieron a la consulta de dermatología del Hospital Pediátrico de Centro Habana en el año 2010, con signos de lesiones pigmentarias de la piel con riesgo de transformación maligna. Se aplicó un instrumento de recogida de datos, con previo consentimiento informado. Se utilizó la estadística descriptiva para el procesamiento de los datos. Resultados: en la población de estudio predominó la adolescencia temprana, el grupo de riesgo con daño actínico crónico y el fototipo III. Estuvo presente el antecedente familiar de cáncer de piel no-melanoma. La lesión más frecuente fue la de nevo melanocítico displásico. Además, se reportó un caso de carcinoma basal en paciente de fototipo II, perteneciente al grupo de riesgo de xeroderma pigmentoso. Conclusiones: se evidenció la transformación maligna de lesiones pigmentarias en piel de pacientes de grupos de riesgo.

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)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Remote Reactor Monitoring

Energy Technology Data Exchange (ETDEWEB)

Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-10-21

The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

2001-07-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

disegnare con ... Alberto Pratelli

Directory of Open Access Journals (Sweden)

Roberto Mingucci

2017-12-01

Full Text Available Con questa breve intervista ad Alberto Pratelli, (non a caso scelto per aprire questa nuova rubrica intendia-mo inaugurare un dialogo con personalità significati-ve del Disegno di Architettura, che consenta riflessioni dedicate alle sue varie dimensioni, oggi più che mai da approfondire. La suggestione a farlo, viene da un’idea di Pablo Rodri-guez Navarro ed abbiamo quindi pensato di avviarla proprio in questo numero, che Pablo ha accettato di curare su un tema a lui particolarmente caro.

Canada-India Reactor (CIR)

Energy Technology Data Exchange (ETDEWEB)

None

1960-12-15

Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

Depresion en pacientes con alteraciones del tiroides

OpenAIRE

Radanovic-Grguric´, Ljiljana; Filakovic´, Pavo; Barkic´, Jelena; Mandic´, Nikola; Karner, Ivan; Smoje, Juraj

2003-01-01

Nuestro estudio fue realizado en un grupo de 53 mujeres con disfunción tiroidea y 28 mujeres con depresión mayor. Empleamos la Escala de la Depresión de Hamilton, la Escala de Autoevaluación de la Depresión de Zung y la Escala sobre la Impresión Clínica Global. Los resultados del estudio demuestran que la mayoría de los pacientes con disfunción tiroidea se mostraron clínicamente significativos en cuanto al trastorno depresivo. Los episodios depresivos son más frecuentes en pacientes con hipot...

REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

International Nuclear Information System (INIS)

Nelson, W.R.

1982-01-01

REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

One piece reactor removal

International Nuclear Information System (INIS)

Chia, Wei-Min; Wang, Song-Feng

1993-01-01

The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

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.)

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.)

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Enfermedad coronaria en pacientes con psoriasis

Directory of Open Access Journals (Sweden)

Walter Masson

2013-10-01

Full Text Available Comunicaciones previas asociaron la psoriasis con la enfermedad coronaria. Desconocemos si en nuestro país o región existe dicha asociación. Se realizó un estudio transversal analizando los datos de la historia clínica electrónica de un sistema de salud de Buenos Aires. Analizamos todos los pacientes mayores de 18 años con diagnóstico de psoriasis entre el 1 de enero de 2003 y el 31 de julio de 2011 y los comparamos con un grupo control, en una relación 2:1, obtenido en forma aleatoria del mismo sistema de salud, apareados por edad y sexo. Determinamos la prevalencia de los factores de riesgo cardiovascular y de enfermedad coronaria. Analizamos la asociación entre la enfermedad coronaria y la psoriasis mediante análisis uni y multivariado. Se incluyeron 3 833 sujetos (1 286 pacientes con psoriasis y 2 547 controles. La prevalencia de hipertensión arterial (50% vs. 38%, p < 0.001, tabaquismo (25% vs. 17%, p < 0.001, diabetes (12% vs. 8%, p < 0.001 y enfermedad coronaria (4.98% vs. 3.06%, p = 0.003 fue mayor en los sujetos con psoriasis en comparación con el grupo control. Independientemente de la edad, la presencia de diabetes, hipertensión arterial o tabaquismo, hubo una asociación significativa entre la enfermedad coronaria y la psoriasis (OR 1.48, IC95% 1.04-2.11, p = 0.03. En conclusión, en esta población de Buenos Aires, los pacientes con psoriasis tuvieron una mayor prevalencia de diabetes, hipertensión arterial, tabaquismo y enfermedad coronaria. La asociación entre la psoriasis y la enfermedad coronaria fue independiente de los factores de riesgo explorados.

Mincle Signaling Promotes Con-A Hepatitis

Science.gov (United States)

Greco, Stephanie H.; Torres-Hernandez, Alejandro; Kalabin, Aleksandr; Whiteman, Clint; Rokosh, Rae; Ravirala, Sushma; Ochi, Atsuo; Gutierrez, Johana; Salyana, Muhammad Atif; Mani, Vishnu R.; Nagaraj, Savitha V.; Deutsch, Michael; Seifert, Lena; Daley, Donnele; Barilla, Rocky; Hundeyin, Mautin; Nikifrov, Yuriy; Tejada, Karla; Gelb, Bruce E.; Katz, Steven C.; Miller, George

2016-01-01

Concanavalin-A (Con-A) hepatitis is regarded as a T cell-mediated model of acute liver injury. Mincle is a C-type lectin receptor (CLR) that is critical in the immune response to mycobacteria and fungi, but does not have a well-defined role in pre-clinical models of non-pathogen mediated inflammation. Since Mincle can ligate the cell death ligand SAP130, we postulated that Mincle signaling drives intrahepatic inflammation and liver injury in Con-A hepatitis. Acute liver injury was assessed in the murine Con-A hepatitis model using C57BL/6, Mincle−/−, and Dectin-1−/− mice. The role of C/EBPβ and HIF-1α signaling was assessed using selective inhibitors. We found that Mincle was highly expressed in hepatic innate inflammatory cells and endothelial cells in both mice and humans. Furthermore, sterile Mincle ligands and Mincle signaling intermediates were increased in the murine liver in Con-A hepatitis. Most significantly, Mincle deletion or blockade protected against Con-A hepatitis whereas Mincle ligation exacerbated disease. Bone marrow chimeric and adoptive transfer experiments suggested that Mincle signaling in infiltrating myeloid cells dictates disease phenotype. Conversely, signaling via other CLRs did not alter disease course. Mechanistically, we found that Mincle blockade decreased the NF-κβ related signaling intermediates, C/EBPβ and HIF-1α, both of which are necessary in macrophage-mediated inflammatory responses. Accordingly, Mincle deletion lowered production of nitrites in Con-A hepatitis and inhibition of both C/EBPβ and HIF1-α reduced the severity of liver disease. Our work implicates a novel innate immune driver of Con-A hepatitis and, more broadly, suggests a potential role for Mincle in diseases governed by sterile inflammation. PMID:27559045

Mincle Signaling Promotes Con A Hepatitis.

Science.gov (United States)

Greco, Stephanie H; Torres-Hernandez, Alejandro; Kalabin, Aleksandr; Whiteman, Clint; Rokosh, Rae; Ravirala, Sushma; Ochi, Atsuo; Gutierrez, Johana; Salyana, Muhammad Atif; Mani, Vishnu R; Nagaraj, Savitha V; Deutsch, Michael; Seifert, Lena; Daley, Donnele; Barilla, Rocky; Hundeyin, Mautin; Nikifrov, Yuriy; Tejada, Karla; Gelb, Bruce E; Katz, Steven C; Miller, George

2016-10-01

Con A hepatitis is regarded as a T cell-mediated model of acute liver injury. Mincle is a C-type lectin receptor that is critical in the immune response to mycobacteria and fungi but does not have a well-defined role in preclinical models of non-pathogen-mediated inflammation. Because Mincle can ligate the cell death ligand SAP130, we postulated that Mincle signaling drives intrahepatic inflammation and liver injury in Con A hepatitis. Acute liver injury was assessed in the murine Con A hepatitis model using C57BL/6, Mincle(-/-), and Dectin-1(-/-) mice. The role of C/EBPβ and hypoxia-inducible factor-1α (HIF-1α) signaling was assessed using selective inhibitors. We found that Mincle was highly expressed in hepatic innate inflammatory cells and endothelial cells in both mice and humans. Furthermore, sterile Mincle ligands and Mincle signaling intermediates were increased in the murine liver in Con A hepatitis. Most significantly, Mincle deletion or blockade protected against Con A hepatitis, whereas Mincle ligation exacerbated disease. Bone marrow chimeric and adoptive transfer experiments suggested that Mincle signaling in infiltrating myeloid cells dictates disease phenotype. Conversely, signaling via other C-type lectin receptors did not alter disease course. Mechanistically, we found that Mincle blockade decreased the NF-κβ-related signaling intermediates C/EBPβ and HIF-1α, both of which are necessary in macrophage-mediated inflammatory responses. Accordingly, Mincle deletion lowered production of nitrites in Con A hepatitis and inhibition of both C/EBPβ and HIF-1α reduced the severity of liver disease. Our work implicates a novel innate immune driver of Con A hepatitis and, more broadly, suggests a potential role for Mincle in diseases governed by sterile inflammation. Copyright © 2016 by The American Association of Immunologists, Inc.

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

2000-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1998-01-01

Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1996-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

Cementos con cenizas volantes

Directory of Open Access Journals (Sweden)

Ossa M., Mauricio

1984-03-01

additions of 20 and 30% .

Casi la generalidad de los estudios realizados sobre cementos con adición de cenizas volantes se refieren a sus características y comportamiento en pastas, morteros y hormigones, siempre en relación con aquéllos del cemento portland. Esta vez, se desarrolló un trabajo experimental orientado a relacionar entre sí los cementos con adiciones de cenizas volantes y de puzolana natural. Para ello se fabricaron a escala de laboratorio cementos de ambos tipos, empleando como materias primas comunes clinker y yeso y, como variables, diferentes porcentajes de las dos adiciones, que cumplieron previamente los requisitos normalizados en cuanto a sus actividades puzolánicas. La calidad de los cementos fabricados resultó adecuada y concordante con la del cemento portland-puzolánico obtenido a escala industrial con los mismos clinker, yeso y puzolana natural de este estudio. Posteriormente, se determinaron las características de los cementos experimentales y se confeccionaron morteros normales para la realización de ensayos físicos y mecánicos. Los resultados de ensayos indicaron que los cementos con adición de cenizas volantes (CCV requieren menos agua para consistencia normal, presentan tiempos de fraguado mayores y expansiones en autoclave menores que los cementos con adición de puzolana (CP. Los calores de hidratación a 7 y 28 días de edad fueron aproximadamente similares para ambos tipos de cemento. En morteros normales, los cementos CCV mostraron menor retracción de secado, mayor retentividad y mayor fluidez (para igual cantidad de agua que los cementos CP. En los ensayos de exudación se observó que ésta depende más de la finura que el tipo de adición. Finalmente, los ensayos mecánicos señalaron que las resistencias a compresión y flexotracción de los morteros con cementos CCV son menores a edades inferiores que 14 días (del orden de 5 a 10% a un día de edad, pero que a partir de entonces pasan a ser mayores que las de

Enzymatic oxidation of volatile malodorous organosulfur compounds in a two-phase reactor

Directory of Open Access Journals (Sweden)

Julio C. Cruz

2015-01-01

Full Text Available En este trabajo reportamos la oxidación de una serie de compuestos organoazufrados volátiles (COV catalizada por la enzima cloroperoxidasa obtenida del hongo Caldariomyces fumago . Los COV se consideran contaminantes atmosféricos debido a su olor desagradable y a su bajo umbral de detección. El sulfuro de etilo, disulfuro de dimetilo, propanotiol, butanotiol y hexanotiol fueron transformados por la enzima en un medio de reacción acuoso a pH 6 y en presencia de peróxido de hidrógeno. El análisis de los productos demostró que los sulfuros fueron oxidados a sus respectivos sulfóxidos, mientras que los tioles fueron oxidados a sus correspondientes disulfuros. Los productos identificados tienen una presión de vapor significativamente menor que los compuestos originales, por lo que son mucho menos volátiles y por tanto no se consideran contaminantes atmosféricos. Se ensambló un reactor de dos fases de 70 mL de volumen con el fin de determinar la eficiencia del tratamiento enzimático. La fase líquida, compuesta por 85% de amortiguador y 15% de solvente orgánico, se puso en contacto con la fase gaseosa, compuesta por aire enriquecido con el sustrato. Usando disulfuro de metilo como sustrato modelo, encontramos únicamente reacción enzimática en este sistema; al controlar la concentración de enzima y de peróxido en la fase líquida se logró transferir el sustrato a la fase acuosa en donde 1 mol de enzima convirtió aproximadamente 12,400 moles de sustrato, resaltando el potencial de los tratamientos enzimáticos para las corrientes gaseosas con mal olor por COV.

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

FBR type reactor core

International Nuclear Information System (INIS)

Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

1995-01-01

Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

Efecto de un tratamiento intensivo con insulina sobre la reactividad plaquetaria en pacientes con hiperglucemia que ingresan con un síndrome coronario agudo

OpenAIRE

Vivas Balcones, Luis David

2011-01-01

El objetivo primario de este trabajo ha sido la valoración del efecto que produce el control estricto de la glucemia mediante un tratamiento intensivo y precoz con insulina sobre la reactividad plaquetaria durante la fase hospitalaria en pacientes que ingresan en una unidad de cuidados intensivos cardiológicos con un síndrome coronario agudo e hiperglucemia, respecto a un tratamiento convencional.

Reactor operation method

International Nuclear Information System (INIS)

Osumi, Katsumi; Miki, Minoru.

1979-01-01

Purpose: To prevent stress corrosion cracks by decreasing the dissolved oxygen and hydrogen peroxide concentrations in the coolants within a reactor container upon transient operation such as at the start-up or shutdown of bwr type reactors. Method: After a condensate has been evacuated, deaeration operation is conducted while opening a main steam drain line, as well as a main steam separation valve and a by-pass valve in a turbine by-pass line connecting the main steam line and the condenser without by way of a turbine, and the reactor is started-up by the extraction of control rods after the concentration of dissolved oxygen in the cooling water within a pressure vessel has been decreased below a predetermined value. Nuclear heating is started after the reactor water has been increased to about 150 0 C by pump heating after the end of the deaeration operation for preventing the concentration of hydrogen peroxide and oxygen in the reactor water from temporarily increasing immediately after the start-up. The corrosive atmosphere in the reactor vessel can thus be moderated. (Horiuchi, T.)

Nuclear reactor buildings

International Nuclear Information System (INIS)

Nagashima, Shoji; Kato, Ryoichi.

1985-01-01

Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

The reactor Cabri

International Nuclear Information System (INIS)

Ailloud, J.; Millot, J.P.

1964-01-01

It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m 3 /h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

Hiperalgesia asociada al tratamiento con opioides

OpenAIRE

A. Gil Martín; M. Moreno García; J. Sánchez-Rubio Ferrández; T. Molina García

2014-01-01

La hiperalgesia inducida por opioides es una reacción paradójica caracterizada por una percepción intensificada de dolor relacionada con el uso de estos medicamentos en ausencia de progresión de la enfermedad o de síndrome de retirada. A diferencia de los casos de tolerancia, definida como pérdida de potencia analgésica durante el uso prolongado de opioides, no se produce mejoría con el escalado de dosis. La hiperalgesia inducida por opioides se ha manifestado en pacientes con dosis de manten...

Paciente con tumor de cuerpo carotideo

Directory of Open Access Journals (Sweden)

Mariuska Forteza Sáez

Full Text Available Los tumores de cuerpo carotideo (paragangliomas son neoplasias altamente vascularizadas, muy poco frecuentes y generalmente benignas, originadas en los quimiorreceptores del cuerpo carotideo. Se presenta el caso de un paciente de 54 años, con aumento de volumen cervical derecho, asintomático, con estudio preoperatorio y angiografía realizados por tomografía axial computarizada, que resultan compatibles con tumor de cuerpo carotideo. Se realiza disección subadventicial, informando la biopsia paraganglioma. El tumor fue completamente resecado, sin evidencia de recurrencia y sin complicaciones.

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J.

1982-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

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

Innovative hybrid biological reactors using membranes; Reactores biologico hibrido innovadores utilizando membranas

Energy Technology Data Exchange (ETDEWEB)

Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

2011-07-01

In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

International Nuclear Information System (INIS)

Joo, Young Sang; Lee, Jae Han

2007-02-01

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Lee, Jae Han

2007-02-15

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

The Maple reactor project

International Nuclear Information System (INIS)

Malkoske, G.R.; Labrie, J.-P.

2003-01-01

MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Methanogenesis in Thermophilic Biogas Reactors

DEFF Research Database (Denmark)

Ahring, Birgitte Kiær

1995-01-01

Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process as ....... Experiments using biogas reactors fed with cow manure showed that the same biogas yield found at 550 C could be obtained at 610 C after a long adaptation period. However, propionate degradation was inhibited by increasing the temperature.......Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...

A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

Future view of total energy system and reactor engineering and reactor physics

International Nuclear Information System (INIS)

Ozawa, T.

1974-01-01

This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

International Nuclear Information System (INIS)

Bacher, P.

1997-01-01

A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

Research reactor support

International Nuclear Information System (INIS)

2005-01-01

Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

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

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.)

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

Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

International Nuclear Information System (INIS)

1975-06-01

The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

2012 review of French research reactors

International Nuclear Information System (INIS)

Estrade, Jerome

2013-01-01

Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

Upgradation of Apsara reactor

International Nuclear Information System (INIS)

Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

2009-01-01

Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

Beneficios de la terapia con resistómetro inspiratorio en los pacientes con esclerosis multiple frente a terapias respiratorias convencionales

OpenAIRE

Martín Sánchez, Carlos

2016-01-01

[ES] La Universidad de Salamanca (USAL) a través del programa de Doctorado en Oncología Clínica realiza un estudio con ASPRODES, la Asociación Salmantina de Esclerosis Múltiple (ASDEM) y la Asociación Zamorana de Esclerosis Múltiple (AZDEM) para llevar a cabo un estudio experimental con el fin de analizar las ventajas que la terapia respiratoria con resistómetro inspiratorio puede aportar a los pacientes con Esclerosis Múltiple frente a las técnicas de fisioterapia respiratoria convencional. ...

Los adolescentes españoles y sus familias: Calidad en la comunicación con el padre y con la madre y conductas de riesgo relacionadas con el consumo de sustancias adictivas

OpenAIRE

Moreno Rodríguez, María del Carmen; Muñoz Tinoco, María Victoria; Pérez Moreno, Pedro Juan; Sánchez Queija, María Inmaculada

2006-01-01

Se presenta una reflexión acerca de cómo transcurren los años de la adolescencia en el contexto de la familia, analizando las transformaciones que se producen en las relaciones con los progenitores. El trabajo empírico se centra en una de las dimensiones que definen la calidad de las relaciones en el contexto familiar -la facilidad en la comunicación entre el adolescente y sus padres- y en su relación con algunas conductas de los adolescentes relacionadas con la salud, como es el caso del con...

Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

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

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

Materials for passively safe reactors

International Nuclear Information System (INIS)

Simnad, T.

1993-01-01

Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

Santiago, una ciudad con temor

Directory of Open Access Journals (Sweden)

Enrique Oviedo S.

1999-04-01

Full Text Available El objetivo general de este artículo es evaluar los efectos de la inseguridad ciudadana en el uso del espacio público. Dicha evaluación exige analizar dos relaciones que se establecen en el ámbito de la violencia: la relación entre victimización y percepción de inseguridad; y la que se establece entre actitudes sociales y resolución pacífica de conflictos nacionales. Para ello, se analizaron las variables victimización, percepción de inseguridad, uso del espacio físico, actitudes hacia el sistema institucional político y social y hacia la resolución de conflictos nacionales, y las posibles relaciones entre ellas. Los datos para realizar el estudio se obtuvieron por medio de una encuesta que se llevó a cabo con 1 200 personas de 18 y 70 años de edad residentes en la ciudad de Santiago. Los resultados indican que Santiago es una ciudad de habitantes con temor y que el aumento de la percepción de inseguridad de sus habitantes contrasta con el hecho de que las tasas de victimización se hayan mantenido, más o menos, constantes en los años que precedieron a la encuesta. El temor se relaciona con el abandono del espacio público físico y sociopolítico, así como con el refugio en los espacios y la vida privados. La actitud de resolver los conflictos por medios no pacíficos es frecuente y se asocia en mayor medida con la inseguridad, la actitud negativa hacia la democracia y la falta de expectativas sobre el futuro del país. Los resultados de este estudio respaldan la idea de que para superar el temor la gente tiende a adaptarse a la realidad adoptando una postura conformista, homogeneizando las creencias y los comportamientos, y sobreestimando la fuerza como medio para resolver las diferencias.

Santiago, una ciudad con temor

Directory of Open Access Journals (Sweden)

Oviedo S. Enrique

1999-01-01

Full Text Available El objetivo general de este artículo es evaluar los efectos de la inseguridad ciudadana en el uso del espacio público. Dicha evaluación exige analizar dos relaciones que se establecen en el ámbito de la violencia: la relación entre victimización y percepción de inseguridad; y la que se establece entre actitudes sociales y resolución pacífica de conflictos nacionales. Para ello, se analizaron las variables victimización, percepción de inseguridad, uso del espacio físico, actitudes hacia el sistema institucional político y social y hacia la resolución de conflictos nacionales, y las posibles relaciones entre ellas. Los datos para realizar el estudio se obtuvieron por medio de una encuesta que se llevó a cabo con 1 200 personas de 18 y 70 años de edad residentes en la ciudad de Santiago. Los resultados indican que Santiago es una ciudad de habitantes con temor y que el aumento de la percepción de inseguridad de sus habitantes contrasta con el hecho de que las tasas de victimización se hayan mantenido, más o menos, constantes en los años que precedieron a la encuesta. El temor se relaciona con el abandono del espacio público físico y sociopolítico, así como con el refugio en los espacios y la vida privados. La actitud de resolver los conflictos por medios no pacíficos es frecuente y se asocia en mayor medida con la inseguridad, la actitud negativa hacia la democracia y la falta de expectativas sobre el futuro del país. Los resultados de este estudio respaldan la idea de que para superar el temor la gente tiende a adaptarse a la realidad adoptando una postura conformista, homogeneizando las creencias y los comportamientos, y sobreestimando la fuerza como medio para resolver las diferencias.

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

The fast breeder reactor

International Nuclear Information System (INIS)

Davis, D.A.; Baker, M.A.W.; Hall, R.S.

1990-01-01

Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

Method of reactor operation

International Nuclear Information System (INIS)

Maeda, Katsuji.

1982-01-01

Purpose: To prevent stress corrosion cracks in stainless steels caused from hydrogen peroxide in reactor operation in which the density of hydrogen peroxide in the reactor water is controlled upon reactor start-up. Method: A heat exchanger equipped with a heat source for applying external heat is disposed into the recycling system for reactor coolants. Upon reactor start-up, the coolants are heated by the heat exchanger till arriving at a temperature at which the dissolving rate is faster than the forming rate of hydrogen peroxide in the coolants, and nuclear heating is started after reaching the above temperature. The temperature of the reactor water is increased in such a manner and, when it arrives at 140 0 C, extraction of control elements is started and the heat source for the heat exchanger is interrupted simultaneously. In this way spikes in the density of hydrogen peroxide are suppressed upon reactor start-up to thereby decrease the stress corrosion cracks in stainless steels. (Horiuchi, T.)

Reactor water sampling device

International Nuclear Information System (INIS)

Sakamaki, Kazuo.

1992-01-01

The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

Reactor power monitoring device

International Nuclear Information System (INIS)

Dogen, Ayumi; Ozawa, Michihiro.

1983-01-01

Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1975-11-01

Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

Reactor materials research as an effective instrument of nuclear reactor perfection