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Sample records for agua hirviente bwr

  1. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  2. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  3. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors; Tendencias recientes en la mitigacion del IGSCC mediante modificaciones en la quimica del agua de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Diaz S, A.; Robles, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  5. Agua(s) remisión de V. nivelador de (las) agua(s).

    OpenAIRE

    2011-01-01

    [ES] Definición del término Agua(s) remisión de V. nivelador de (las) agua(s). en el diccionario Dicter. [EN] Definition of the word Agua(s) remisión de V. nivelador de (las) agua(s). in the dictionary Dicter.

  6. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  7. Sector agua

    OpenAIRE

    Mabel Scharfhausen

    2012-01-01

    El agua constituye un recurso esencial para muchos aspectos de la vida humana como son el desarrollo y el bienestar humano. Gracias a los importantes avances en TICs es posible avanzar hacia soluciones que ofrecen una gestión eficiente e integral de los recursos hídricos, proporcionando datos en tiempo real respecto a la calidad, al consumo y la demanda del agua que permiten. Esto, no solamente para tener un mejor control del recurso, sino para hacer seguimiento del mismo y, quizás lo más imp...

  8. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

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

  9. MAAP BWR application guidelines

    International Nuclear Information System (INIS)

    The MAAP Thermal-Hydraulic Qualification and Application Project has as its objective to identify those thermal-hydraulic phenomena modeled in MAAP which are important in predicting severe accident sequences, to qualify those models and to provide guidelines for use of the code. This report provides user guidelines for use of the BWR version of MAAP. The report includes a discussion of the important features of the BWR that are modeled in MAAP, the MAAP modeling of phenomena important to predicting severe accidents and user guidelines for several accident sequences

  10. BWR internal cracking issues

    International Nuclear Information System (INIS)

    The regulatory issues associated with cracking of boiling water reactor (BWR) internals is being addressed by the Nuclear Regulatory Commission (NRC) staff and is the subject of a voluntary industry initiative. The lessons learned from this effort will be applied to pressurized water reactor (PWR) internals cracking issues

  11. Development of advanced BWR

    International Nuclear Information System (INIS)

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

  12. Trends in BWR transient analysis

    International Nuclear Information System (INIS)

    While boiling water reactor (BWR) analysis methods for transient and loss of coolant accident analysis are well established, refinements and improvements continue to be made. This evolution of BWR analysis methods is driven by the new applications. This paper discusses some examples of these trends, specifically, time domain stability analysis and analysis of the simplified BWR (SBWR), General Electric's design approach involving a shift from active to passive safety systems and the elimination/simplification of systems for improved operation and maintenance

  13. BWR stability analysis

    International Nuclear Information System (INIS)

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

  14. Modernising Sweden's oldest BWR

    International Nuclear Information System (INIS)

    A new range of digital, programmable process control systems has been used to upgrade the instrumentation and control room of Oskarshamn 1's power station. It will enable Sweden's oldest BWR reactor to comply with higher safety levels, improving the operator's overview of production, and reducing the risk of human error. The ''Advant Power'' range of systems will also be used for future planned improvements to the control room. (UK)

  15. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

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

  16. BWR AXIAL PROFILE

    Energy Technology Data Exchange (ETDEWEB)

    J. Huffer

    2004-09-28

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

  17. Diagnostico integrado de agua

    Directory of Open Access Journals (Sweden)

    Yanely de la Caridad Esquijarosa Abradelo

    2010-01-01

    Full Text Available Se aborda una metodología para el diagnóstico integrado del agua, basado en herramientas que brindan la información necesaria, teniendo en cuenta la necesidad de poner en práctica nuevas vías que posibiliten los índices cualitativos y cuantitativos de los servicios de agua, de acuerdo a la política energética y medioambiental y a las prioridades que demandan los objetivos sociales y económicos de cada territorio, que conllevan a la implantación de un sistema de gestión integral del agua.

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

    International Nuclear Information System (INIS)

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

  19. BWR fuel performance

    International Nuclear Information System (INIS)

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

  20. Concesiones de aguas

    Directory of Open Access Journals (Sweden)

    Carlos Labaure Aliseris

    2014-07-01

    Full Text Available El objeto del estudio que nos proponemos realizar es el de las concesiones de aguas.Dicho tema se inscribe en el más amplio de las concesiones en general, y de las concesiones de bienes del dominio público en particular.El estudio especial de las concesiones de aguas se justifica, ya que como en el caso de las demás concesiones sobre bienes del dominio público y privado del Estado, presenta particularidades en su régimen jurídico.El tema planteado fue objeto de nuestro análisis hace unos años, ya que precisamente nuestra tesis para aspirar al título de Profesor Adscripto fue el Código de Aguas, en que uno de los capítulos era el "Uso de las Aguas", que incluía a los permisos y concesiones de uso

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  2. BWR type reactor

    International Nuclear Information System (INIS)

    In a coolant circulation in BWR type reactors, since the mixed stream of steam fluid undergoes a great resistance, the pressure loss due to the flow rate distribution when the coolants flow from the upper plenum into the stand pipe is increased upon passing stand pipe. Also in the spontaneous recycling reactor, pressure loss is still left upon passing the swirling blade of a gas-liquid separator. In view of the above, a plurality of vertical members each having a lower end opened to a gas-liquid two phase boundary and an upper end directly suspended from a steam dryer to the gas-liquid separator. The liquid droplets from the 2-phase boundary heated in the reactor core and formed into a mixed gas-liquid 2-phase stream is directed in the vertical direction accompanied with the steam. The liquid droplets spontaneously fallen by gravity from greater ones successively and the droplets in the steam abutted against the vertical member are fallen as a liquid membrane. Thus, the gas-liquid separation is conducted, the dry steam is directly flown into the steam dryer, thereby capable of providing a gas-liquid separator having gas-liquid separation performance with lower loss than usual. (N.H.)

  3. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  4. Potencial del agua del suelo

    OpenAIRE

    Bustamante Heliodoro

    2012-01-01

    La energía potencial del agua presenta diferencias de un punto del suelo a otro; esas diferencias son las que originan el movimiento del agua de acuerdo a la tendencia universal de la materia en el sentido de moverse de donde la energía potencial es mayor a donde dicha energía es menor. En el suelo el agua en consecuencia se mueve hacia donde su energía decrece hasta lograr su estado de equilibrio. Se desprende entonces que la cantidad de energía potencial absoluta contenida en el agua, no es...

  5. Conocemos el agua experimentando

    OpenAIRE

    Somavilla Bouzán, Sandra

    2013-01-01

    En este Trabajo Fin de Grado (TFG) se quiere poner de manifiesto la importancia de las ciencias y la experimentación durante las primeras etapas educativas. Para conocer el entorno inmediato se han diseñado una serie de actividades experimentales relacionadas con el conocimiento de una de las materias más importantes para la vida, el agua. Dichas actividades, recogidas bajo el nombre “Conocemos el agua”, se han llevado a la práctica en un aula real de Educación Infantil con niños/as de 5-6 añ...

  6. Improvement for BWR operator training

    International Nuclear Information System (INIS)

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

  7. BWR radiation control: plant demonstration

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  8. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B4C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  9. El agua: un recurso esencial

    Directory of Open Access Journals (Sweden)

    Alicia Fernández Cirelli

    2012-01-01

    Full Text Available El agua es esencial para la vida y para el desarrollo de las sociedades. Posee propiedades únicas. En este trabajo se describen: la composición de las aguas naturales; la calidad de agua para los diferentes usos y su deterioro y los parámetros físico-químicos y biológicos usados como indicadores de calidad. Por último, se resalta la importancia de la representatividad del muestreo y la calidad de las mediciones químicas para que los resultados obtenidos permitan conclusiones válidas.

  10. BWR 90 - the advanced BWR of the 1990s

    International Nuclear Information System (INIS)

    The future global role of nuclear power will be determined by its ability to provide economical and safe energy. Nuclear power, like any other substantial contributor to the world's energy needs, must be generated at an acceptable cost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a 'cautious evolution'; for the next decade the company will largely base its offerings to the market on its 'evolutionary' light water reactor design, the BWR 90. This design builds closely on the experience from successful construction and operation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average load factor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very best performing plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR design development focuses on meeting requirements from utilities as well as new regulatory requirements. A particular emphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimization of buildings and containment design to decrease construction time and costs, and selection of materials as well as maintenance and operating procedures to even further reduce occupational radiation exposures. (orig.)

  11. Long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Considering next generation requirement for nuclear plants, a long cycle operating simplified BWR (LSBWR) concept is proposed. The major features of LSBWR are; 1) Long cycle operation core using uranium fuels; 2) Simplified system and component as well as passive systems; 3) Combined building concept with ship hull structure. This concept have potential to reduce construction cost and to Increase availability. Safety feature of LSBWR makes possible to attain no evacuation capability in case of a severe accident. Further research and development is underway. (author)

  12. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

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

  13. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

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

  14. Ciclo cósmico del agua

    OpenAIRE

    Martínez-Frías, J.

    2008-01-01

    1 página.-- Trabajo presentado en la Exposición Internacional "Agua y Desarrollo Sostenible" celebrada en Zaragoza del 14 de Junio al 14 de septiembre del 2008 (Pabellón: Torre del Agua. Agua para la vida).

  15. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

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

  16. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-22

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

  17. 44 BWR Waste Package Loading Curve Evaluation

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU

  18. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  19. a considerar. II. Aguas residuales urbanas

    Directory of Open Access Journals (Sweden)

    Fernando Ortega Sastriques

    2007-01-01

    Full Text Available El precio del agua potable es elevado y en muchas ciudades su disponibilidad está limitada y disminuirá en el futuro. La agricultura urbana es una concurrente en el uso del agua, en el futuro es muy probable que haya fuertes limitaciones para el riego con agua potable. Así es que para la producción de vegetales en medio urbano será necesario el empleo de agua de fuentes alternativas y no de los acueductos. Por lo general, el agua del subsuelo en el entorno de las ciudades, así como las corrientes superficiales está contaminada. En este trabajo se señalan los aspectos a tomar en cuenta para el riego seguro con aguas residuales, incluyendo las aguas albañales.

  20. Agua Caliente and Their Music.

    Science.gov (United States)

    Ryterband, Roman

    1979-01-01

    Discusses the traditional music of the Agua Caliente band of California's Desert Cahuilla Indian tribe, including accompanying instruments, types of songs, thematic material, and performance routines. Exploring the structure of the music, the article describes meter, tempo, harmony and tonal gravitations, and use of words. (DS)

  1. Descubriendo que el agua moja

    OpenAIRE

    Renán Vega Cantor

    2014-01-01

    La influencia de los Estados Unidos en el Ejército colombiano. 1951-1959. Saúl Mauricto Rodríguez Hernández. La Carreta Editores y Universidad Nacional de Colombia, Departamento de Historia, Facullad de Ciencias Humanas, sede Bogotá, Colección Ojo de agua, 2006, 145 págs.

  2. Túnel bajo el agua

    Directory of Open Access Journals (Sweden)

    Stamhuis, E.

    1964-03-01

    Full Text Available This article describes the results obtained by the Dutch Central Institute of Scientific Research when studying the problem of underwater tunnels. This Institute has sought to establish general specifications for the design of this type of tunnels, so that they shall be more efficient than those at present in use. Firstly, a study was made of the strength, waterproof qualities, and most rational type of ventilation, with a view to drawing up of general principles on this aspect of the design. An analysis is given of the shape, suitable cross sections and constructive methods, and these are compared with those adopted in the construction of underwater tunnels at present in use in America. Particular attention has been given to designs which are largely based on prefabrication and the use of factory made structural parts, and also on the transport and erection of these at the site. The author goes on to discuss the interplay of soil and water pressures, buoyancy forces, and cases in which the tunnel is subjected to important bending and shear forces on particular sections of it. In this connection care is taken to give the theoretical considerations a ready practical interpretation. In particular, a description is given of special distance pieces which connect the two tubes making up a tunnel, also the system of joining two sections of the tube, and the measure of watertightness and flexibility which such joints provide. Finally, a summary is given of the theoretical foundations, as well as a number of observations and items of advise, which are applicable to these types of structures.El ingeniero A. Eggink describió, en un trabajo publicado últimamente, un nuevo tipo de túnel bajo el agua para canalizar la circulación rodada. En este trabajo se desarrolla un análisis sobre las consideraciones que han jugado un papel predominante en la redacción del proyecto de ejecución de un túnel subacuo. Acompañan al trabajo los resultados referentes

  3. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

    During the next decade a rise in the energy demand is expected worldwide, and this will in particular call for electricity generation capacity. A number of old generating plants, both nuclear and other plants, will probably have to be shut down for aging reasons, and their replacement will enhance the need for new generating capacity. The ABB Atom considers this situation to be met with a 'cautious evolution'. The offerings will largely be based on 'evolutions' of the successful light water reactor BWR 75. The new, evolutionary plant design of ABB Atom is the BWR 90. It can be designed, licensed and constructed in accordance with any safety regulations now in force or envisaged in the Western world. Emphasis has been, and will be, placed on features that facilitates licensing, shortens construction time and keeps electricity generation costs favourable. ABB also continues to develop a design of the 'passive' type, such as the 'passive' PIUS system, for possible deployment in the future. These efforts are more long-term activities, since development, verification and licensing of distinctly 'new' reactor concepts will have an extensive lead time. This paper presents the BWR 90 and its current status. The design is based on that of its forerunner, the BWR 75 standard design, taking into account the experiences gained from design and engineering, construction, commissioning, and operation of BWR 75 plants, the needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings. (author) 4 figs

  4. ARES - a new BWR simulator

    International Nuclear Information System (INIS)

    Coupling the three-dimensional Analytic Function Expansion Nodal (AFEN) nodal model developed within the READY project in 2000-2001 with a four-equation based stationary-state thermalhydraulics module and a new cross section model, a basis has been created for a sophisticated BWR simulator code. Motivation for distilling the efforts into a new simulator, named ARES (AFEN Reactor Simulator), can be summarized in three main points: Stationary-state analyses required by the safety authorities must be independent from the calculations made by the power utilities. Using a different simulator for some calculations is an effective method for obtaining independent results; In order to keep up with the development in the core analysis field, a 'test bench' is required for testing and evaluating new ideas and models. In addition, accuracy of the commercial codes and the models incorporated in them can be evaluated by benchmarking them against the new simulator; and Writing a new program from scratch is potentially a good way to transfer experience from the first generation of Finnish nuclear engineers. It also gives the opportunity to re-evaluate some of the ideas used in the older codes written in times of significantly smaller computer capacity. (orig.)

  5. Thermal-hydraulics in BWR

    International Nuclear Information System (INIS)

    In the heat transferring flow in BWRs, the heightening of heat transfer performance accompanying the development of new fuel for the purpose of reducing spent fuel generation and the improvement of fuel economy, the heightening of performance and the reduction of size of various heat exchangers, the development of the safety devices, of which the constitution is simple, the reliability is high, and the operation is easy, and so on are expected. As for ABWRs, thermal output is 3926 MW, and electricity output is 1356 MW. The system constitution of ABWR is shown. The main change from BWR to ABWR is the adoption of internal pumps, reinforced concrete containment vessels and electric control rod drive. For evaluating the limit output of high burnup fuel assemblies, the subchannel analysis and the effect that spacers exert to the limit output are explained. The heat transferring flow in moisture separation heater, condenser and feed water heater is reported. The heat transferring flow in passive containment vessel cooling system of water wall type and condensing type is described. (K.I.)

  6. Natural convection type BWR reactor

    International Nuclear Information System (INIS)

    In a natural convection type BWR reactor, a mixed stream of steams and water undergo a great flow resistance. In particular, pressure loss upon passing from an upper plenum to a stand pipe and pressure loss upon passing through rotational blades are great. Then, a steam dryer comprising laminated dome-like perforated plates and a drain pipe for flowing down separated water to a downcomer are disposed above a riser. The coolants heated in the reactor core are boiled, uprise in the riser as a gas-liquid two phase flow containing voids, release steams containing droplets from the surface of the gas-liquid two phase, flow into the steam dryer comprising the perforated plates and are separated into a gas and a liquid. The dried steams flow to a turbine passing through a main steam pipe and the condensated droplets flow down through the drain pipe and the downcomer to the lower portion of the reactor core. In this way, the conventional gas-liquid separator can be saved without lowering the quality of steam drying to reduce the pressure loss and to improve the operation performance. (N.H.)

  7. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  8. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-craking (SCC) susceptibility of Types 304, 316NG, and 347 stainless (SS); (b) fracture-mechanics crack-growth-rate measurements on these materials and weld overlay specimens in different environments; and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on Types 304 and 316NG SS under crevice and non-crevice conditions in 2890C water containing 0.25 ppM dissolved oxygen with low sulfate concentrations indicate that SCC initiates at very low strains (0 in both directions, and then grew at high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones

  9. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    A relational database management system with a query language, Reference 1, has been used to develop a Boiling Water Reactor (BWR) fuel channel tracking system on a microcomputer. The software system developed implements channel vendor and Nuclear Regulatory Commission recommendations for in-core channel movements between reactor operating cycles. A BWR Fuel channel encloses the fuel bundle and is typically fabricated using Ziracoly-4. The channel serves three functions: (1) it provides a barrier to separate two parallel flow paths, one inside the fuel assembly and the other in the bypass region outside the fuel assembly and between channels; (2) it guides the control rod as it moves between fuel assemblies and provides a bearing surface for the blades; and (3) it provides rigidity for the fuel bundle. All of these functions are necessary in typical BWR core designs. Fuel channels are not part of typical Pressurized Water Reactor (PWR) core designs

  10. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

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

  11. Cutting BWR feedwater crud levels further

    International Nuclear Information System (INIS)

    Reducing iron input to the BWR primary system is an important first step in decreasing radiation fields and occupational exposure. For feedwater iron, the specified optimum concentration is 0.1-0.5 ppb, as demonstrated by the newest ''low crud'' plants operating in Japan. Recent advances in condensate filtration will achieve the levels needed for optimised water chemistry and promise great benefits. Of particular interest is a newly developed filterdemineraliser septum that separates the filtration and ion exchange functions to allow each to be specifically optimised for BWR conditions. (author)

  12. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-cracking (SCC) susceptibility of types 304, 316 NG, and 347 stainless steel (SS), (b) fracture-mechanics crack growth rate measurements on these materials and weld overlay specimens in different environments, and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on types 304 and 316 NG SS under crevice and non-crevice conditions in 2890C water containing 0.25 ppm dissolved oxygen with low sulfate concentrations indicate that SCC initiates at low strains (3%) in the nuclear grade material. Crack growth measurements on fracture-mechanics-type specimens, under low-frequency cyclic loading, show that the type 316 NG steel cracks at a somewhat lower rate (≅ 40%) than sensitized type 304 SS in an impurity environment with 0.25 ppm dissolved oxygen; however, the latter material stops cracking when sulfate is removed from the water. Crack growth in both materials ceases under simulated hydrogen-water chemistry conditions (6 ppb oxygen) even with 100 ppb sulfate present in the water. An unexpected results was obtained in the test on a weld overlay specimen in the impurity environment, viz., the crack grew to the overlay interface at a nominal rate, branched at 900 in both directions, and then grew at a high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones. (orig.)

  13. a considerar. I. Aguas salinas o alcalinas

    Directory of Open Access Journals (Sweden)

    Fernando Ortega Sastriques

    2007-01-01

    Full Text Available El precio del agua potable es elevado y en muchas ciudades su disponibilidad está limitada y disminuirá en el futuro. La agricultura urbana es una concurrente en el uso del agua. En el futuro es muy probable que haya fuertes limitaciones para el riego con agua potable. Así es que para la producción de vegetales en medio urbano será necesario el empleo de agua de fuentes alternativas y no de los acueductos. La extracción de agua del manto freático para el riego es una práctica común que se incrementará en el futuro previsible. Casi todas las grandes urbes de Cuba tienen costas, es frecuente que los acuíferos subterráneos de estas ciudades presenten algún contenido de fácil solubilidad; o bien en otras ciudades interiores el agua puede ser alcalina. El riego con estos tipos de agua deprimen los rendimientos y pueden dañar los suelos y substratos. En este artículo se explican los aspectos que deben considerarse para poder emplear esta agua con seguridad.

  14. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials

  15. AGUA TIBIA PRIMITIVE AREA, CALIFORNIA.

    Science.gov (United States)

    Irwin, William P.; Thurber, Horace K.

    1984-01-01

    The Agua Tibia Primitive Area in southwestern California is underlain by igneous and metamorphic rocks that are siilar to those widely exposed throughout much of the Peninsular Ranges. To detect the presence of any concealed mineral deposits, samples of stream sediments were collected along the various creeks that head in the mountain. As an additional aid in evaluating the mineral potential, an aeromagnetic survey was made and interpreted. A search for records of past or existing mining claims within the primitive area was made but none was found. Evidence of deposits of metallic or nonmetallic minerals was not seen during the study.

  16. El abastecimiento de agua a Cartagena

    Directory of Open Access Journals (Sweden)

    Miguel Borja Bernabé Crespo

    2015-01-01

    Full Text Available La escasez de agua en la región del sureste peninsular, en la que se ubica Cartagena y su comarca,ha obligado a buscar y traer recursos foráneos de agua para atender sus necesidades de abastecimiento.Inicialmente, al no existir cursos permanentes de agua, se recurre a pozos y aljibes, apequeños acueductos desde fuentes y manantiales de aforo escaso, y a alumbramientos de subálveasmediante galerías con lumbreras. Las demandas aumentan y se plantea la traída de aguas dela parte alta de la Cuenca del Segura. Mediante las infraestructuras de la Mancomunidad de losCanales del Taibilla (MCT, creada por Decreto Ley de 4 de octubre de 1927 llega el agua hastaCartagena en 1945. En los últimos treinta y cinco años, a estas aguas se han unido caudales delTrasvase Tajo-Segura y finalmente aguas procedentes de desaladoras. La gestión del abastecimientode agua, a lo largo de la historia de Cartagena (desde la fundación de Carthago Nova enel 229 a.C. hasta los primeros años del siglo XXI ha generado un rico patrimonio de infraestructurashidráulicas y una experimentada cultura de usos del agua.Se expone la situación del abastecimiento de Cartagena (ciudad y barrios como demandante derecursos propios y foráneos de agua. La inversión en las redes de distribución y la concienciaciónde la mayor parte de los usuarios en un uso sostenible, que permiten un continuado abastecimientode «agua de boca», así como la regeneración de las aguas residuales para otros usos. Eneste trabajo geográfico (espacial e histórico (diacrónico, las fuentes utilizadas han sido las entrevistasa los gestores de las empresas encargadas de la distribución del agua en alta (conduccionesde gran diámetro como las de la MCT y en baja (conducciones de pequeño diámetro como las deHIDROGEA, y el acceso a la documentación del Archivo Municipal de Cartagena, del Regionalde Murcia, de la Mancomunidad de Canales del Taibilla, y el de la Confederación Hidrográficadel Segura

  17. Process inherent ultimate safety/boiling-water reactor PIUS/BWR

    International Nuclear Information System (INIS)

    This document is a series of viewgraphs on: design basis of PIUS/BWR, definition of PIUS/BWR, mechanisms of safe shutdown and afterheat cooling, advantages of PIUS/BWR, and research and development requirements

  18. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (

  19. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

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

  20. Calidad del agua potable en Huesca

    OpenAIRE

    Sancho Arroyo, Sandra; Pablo Aranda, Miriam

    2010-01-01

    El agua es uno de los recursos naturales fundamentales; es, seguramente el recurso que condiciona de manera prioritaria el desarrollo socioeconómico de los pueblos y la mejora del bienestar de la población. El agua presta un extraordinario servicio a la comunidad para mejorar su calidad de vida, pero, sobre todo, para satisfacer sus necesidades básicas, tales como la alimentación y la salud. Sin agua no son posibles las actividades agropecuarias y, especialmente, las agrícolas, que son el may...

  1. Service experience of BWR pressure vessels

    International Nuclear Information System (INIS)

    The overall service experience with Boiling Water Reactor (BWR) pressure vessels has been excellent. The only significant factor that impacted the service performance has been thermal fatigue cracking of feedwater inlet nozzle. This concern has been mitigated by eliminating the source of thermal cycling stress through design and operational changes. Although stress corrosion cracking has occurred in early atypical steam generator vessel designs, analysis and field experience has indicated that this mechanism is not expected in the BWR reactor pressure vessel (RPV). Other limited materials related cracking problems have been associated with RPV stainless steel and nickel-base alloy attachments and penetrations. Solutions to these problems have involved design and materials modifications. Finally, due to the low end of life fluence resulting from the large core-to-RPV-wall water annulus, irradiation embrittlement effects are minimal

  2. Examination of overlay repaired BWR pipe joints

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) in a large number of austenitic stainless steel girth welds in boiling water reactor (BWR) piping has prompted the development of the weld overlay for repair (WOR) as a short-term remedy. It is necessary to examine the deposited overlay weld material for adequate definition of its condition and to monitor the overlaid IGSCC to determine if it grows past the bounds assumed in the design of the repair. This paper reports on NDE techniques evaluated using weld overlaid pipe samples containing known defects, overlaid samples removed from BWR service, and overlaid weld joints in plant. These samples included overlays containing fabrication defects and overlaid pipes containing deep and shallow laboratory- and service-induced IGSCC

  3. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

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

  4. The design of large natural circulation BWR's

    International Nuclear Information System (INIS)

    Boiling water reactors (BWR) with natural circulation are applied for capacities up to 60 MWe. Based on scale studies, however, it appears that larger production units are more efficient. It is recommended to investigate the bottlenecks in realizing larger reactors (>1000 MWe). The aim of the study on the title subject is to study to what extent the production capacity of BWRs with natural circulation can be increased. Based on data from the literature a simple analytic method has been chosen and existing BWR designs were compared. Capacities of 1300 MWe appear to be possible. These reactors will have a smaller pin diameter and a lower water supply temperature. Also steam separators with a minor pressure reduction must be available. The reliability of the stability measurement must be increased. Based on the results of this investigation the priorities for research on the design of future BWRs have been determined

  5. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

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

  6. Recent developments in BWR water chemistry

    International Nuclear Information System (INIS)

    Water chemistry is of critical importance to the operation and economic viability of the Boiling Water Reactor (BWR). A successful water chemistry program will satisfy the following goals: - Minimize the incidence and growth of SCC/IASCC, - Minimize plant radiation fields controllable by chemistry, -Maintain fuel integrity by minimizing cladding corrosion, - Minimize flow-accelerated corrosion (FAC) in balance-of-plant components. The impact of water chemistry on each of these goals is discussed in more detail in this paper. It should be noted that water chemistry programs also include surveillance and operating limits for other plant water systems (e.g., service water, closed cooling water systems, etc.) but these are out of the scope of this paper. This paper reviews developments in water chemistry guidelines for U.S. BWR nuclear power plants. (author). 2 figs., 2 tabs., 7 refs

  7. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

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

  8. EASY 5 BWR simulation model for digital feedwater control design

    International Nuclear Information System (INIS)

    The development of a BWR simulation model in support of a program to design and evaluate the digital feedwater control system for the Monticello Boiling Water Reactor (BWR) is described. This model was developed in the EASY5 simulation language in conjunction with EPRI's Modular Modeling System (MMS) two-phase Library. The model consists of three main elements: the BWR reactor vessel module, the feedwater system model, and the steamline model. Transient results for the BWR vessel module and the feedwater system model are presented

  9. Swedes repair BWR thermal fatigue cracks

    International Nuclear Information System (INIS)

    The discovery of cracks in the feedwater and shutdown cooling systems of Sweden's Barseback 2 BWR in 1980 led to investigations in other Swedish nuclear power stations. Similar cracks were found and the defective parts repaired or replaced before being returned to service. The cause of the cracks has been evaluated and efforts are being made to prevent a recurrence. Experience with Ringhals 1, Orkarsham 2 and Forsmark 1 systems are also described. (author)

  10. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

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

  11. Advanced Neutronics Tools for BWR Design Calculations

    International Nuclear Information System (INIS)

    This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)

  12. Analysis of BWR lattices to recycle americium

    International Nuclear Information System (INIS)

    This study was carried out to assess the ability to eliminate meaningful quantities of americium in a primarily thermal neutron flux by 'spiking' modern BWR fuel with this minor actinide (MA). The studies carried out so far include the simulation of modern 10 x 10 BWR lattices employing the Westinghouse lattice physics code PHOENIX-4 alongside validation studies using MCNP5 models of the same lattices that were spatially depleted via the MONTEBURNS code coupling to ORIGEN. When considering the total inventory of minor actinides in Am-spiked pins, excluding isotopes of uranium and plutonium, the results indicate that a reduction of approximately 50% or more in the total mass inventory of these minor actinides is viable within the selected pins. Therefore, these preliminary results have encouraged the extension of this work to the development of improved lattice designs to help optimize the transmutation rates as well as absolute MA inventory reductions. The ultimate goal being to design batches of these advanced BWR bundles alongside multi-cycle core reload strategies. (authors)

  13. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

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

  14. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

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

  15. Decontamination techniques for BWR power generation plant

    International Nuclear Information System (INIS)

    The present report describes various techniques used for decontamination in BWR power generation plants. Objectives and requirements for decontamination in BWR power plants are first discussed focusing on reduction in dose, prevention of spread of contamination, cleaning of work environments, exposure of equipment parts for inspection, re-use of decontaminated resources, and standards for decontamination. Then, the report outlines major physical, chemical and electrochemical decontamination techniques generally used in BWR power generation plants. The physical techniques include suction of deposits in tanks, jet cleaning, particle blast cleaning, ultrasonic cleaning, coating with special paints, and flushing cleaning. The chemical decontamination techniques include the use of organic acids etc. for dissolution of oxidized surface layers and treatment of secondary wastes such as liquids released from primary decontamination processes. Other techniques are used for removal of penetrated contaminants, and soft and hard cladding in and on equipment and piping that are in direct contact with radioactive materials used in nuclear power generation plants. (N.K.)

  16. Advanced neutronics tools for BWR design calculations

    International Nuclear Information System (INIS)

    This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy

  17. Reliability innovations for AREVA NP BWR fuel

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to light water reactors worldwide, representing today more than 185,000 fuel assemblies on the world market including more than 63,000 fuel assemblies for boiling water reactors (BWRs). ATRIUM trademark 10 fuel assemblies have been supplied to a total of 32 BWR plants worldwide resulting in an operating experience over 20,250 fuel assemblies. ATRIUM trademark 10XP and ATRIUM trademark 10XM are AREVA NP's most recent fuel assembly designs featuring improved fuel utilization and achieving high margins to operating limits while maintaining very good reliability. Nevertheless, fuel failures are still encountered in all modern and advanced fuel assembly designs leading to significant operating limitations or unplanned shutdowns of nuclear power plants. The majority of fuel failures in BWR plants are caused by debris fretting, with PCI induced failures being a second leading cause. AREVA NP runs programs to study these root causes and to develop product solutions as part of the continuous improvement process within the Zero Tolerance for Failure (ZTF) initiative. The focus of the ZTF initiative is to further upgrade BWR fuel assembly reliability to achieve the goal of failure free fuel. In the following, two major product improvements are described that will significantly contribute to this goal: - Improved FUELGUARD trademark Lower Tie Plate - Chamfered Fuel Pellet Design (orig.)

  18. TARMS2, New-generation BWR core management system

    International Nuclear Information System (INIS)

    Toshiba has developed a prototype of a new-generation core management system for the boiling water reactor (BWR) TARMS2. It contains function modules for core monitoring and core prediction analysis. TARMS2 is equipped with an advanced three-dimensional BWR core physics model, LOGOS

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

    International Nuclear Information System (INIS)

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

  20. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

  1. Local instability in BWR reactor simulator

    International Nuclear Information System (INIS)

    ''Local'' oscillations in thermal-hydraulic quantities have been observed while performing system code calculations related to PIPER-ONE BWR simulator; void fraction oscillations were detected at the channel middle elevation but not at channel bottom and top. This paper deals with an experimental investigation of ''local'' instability starting from experimental data obtained with the PIPER-ONE facility. The considered test is PO-SD-5B; the post-test analysis has been carried out by RELAP5/MOD2 code. In particular the analysis concerns the experimental and calculated oscillations of pressure drop in the core region and the calculated oscillations of channel void fraction

  2. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  3. Development of ECP models for BWR applications

    International Nuclear Information System (INIS)

    The electrochemical corrosion potential (ECP) of stainless steel has been measured under simulated Boiling Water Reactor (BWR) coolant circuit conditions using a rotating cylinder electrode. Based on the results of measurements an empirical model has been developed to predict the ECP of structure materials in a BVTR primary circuit as a function of H2, O2, and H2O2 concentrations in reactor coolant and water flow velocity. The ECP modeling results using the H2, O2, and H2O2 concentrations calculated by the radiolysis model are compared with the available reactor internal ECP data obtained in an operating reactor

  4. Fracture assessment of a BWR pump nozzle

    International Nuclear Information System (INIS)

    Fracture mechanics calculations are performed to support the non-destructive testing (NDT) qualification programs for pump nozzle investigations of boiling water reactor (BWR) nozzles of reactor pressure vessels (RPVs), with the aim of the determination of qualification defects, which are located in the Inconel 182 weld of the pump nozzle at the bottom of the RPV. The ferritic nozzle and housing have an Inconel buttering and each part is cladded with Inconel 182 before it is mounted. All theses weldments are heat treated after welding; only the connecting weldment between pump housing and nozzle, which is also an Inconel 182 weld, performed on site, is in the as welded condition. (author)

  5. La calidad de las aguas españolas

    OpenAIRE

    Álvarez Cobelas, Miguel; Cabrera, Francisco, (S.I.) (1724-1799)

    1997-01-01

    La calidad del agua es un concepto de ciencia aplicada, variable en el espacio, en el tiempo y según la utilización del agua, que se suele plasmar en una normativa legal. En España la calidad de las aguas continentales ha sufrido un empobrecimiento en las últimas décadas, lo cual, ha repercutido sobre la vida en ellos. O dicho de otro modo, la calidad de la vida depende de la calidad del agua.

  6. en el riego con aguas residuales

    Directory of Open Access Journals (Sweden)

    Reynaldo Roque R.

    2006-01-01

    Full Text Available Es una realidad mundial, actual, que el agua potable es cada día más escasa para todos los usos en las actividades diarias de la población. La agricultura es el mayor consumidor de agua, con marcado incremento en los países en vías de desarrollo donde a su vez se requiere de una mayor producción de alimentos. En áreas urbanas y periurbanas el agua que se requiere para los cultivos hortícolas y riego de campos deportivos y de jardines compite fuertemente por las necesidades crecientes de la población. El uso de aguas residuales de las plantas de tratamiento y lagunas de oxidación se presenta como una alternativa viable para ser utilizada como agua de riego, empleando los sistemas de riego de pivote central, tecnología de punta que ha evolucionado cualitativamente en los últimos años, ofreciendo paquetes integrales, desde los módulos de emisores hasta las posibilidades de programación de los riegos que la hacen grandemente atractivas, técnica y económicamente para estos fines.

  7. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

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

  8. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

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

  9. Dynamic analysis of BWR scram reactivity characteristics

    International Nuclear Information System (INIS)

    An extensive study of BWR scram reactivity behavior is presented. It is based on a space-time analysis of a BWR/4 code using the two-dimensional (R, Z) dynamics code BNL-TWIGL which includes a two-phase thermal-hydraulic model. Calculations were made of the sensitivity of scram to physical quantities such as initial control rod position and power distribution, scram speed, system pressure and varying inlet flow rate and temperature. The end-of-cycle Haling operating condition was found to give rise to the limiting scram reactivity function. Even with scram a power surge was found to be possible with severely decreasing inlet temperature. Calculations were also made to find the effect on scram of commonly used modeling approximations. These included the effect of neglecting delayed neutrons (conservative), using a time invariant void distribution (non-conservative) and defining point kinetics parameters such as reactivity, amplitude function and generation time in terms of different weighting functions. The importance of defining point kinetics parameters consistent with their use in plant transient analyses was demonstrated with particular emphasis on the role of ''residual reactivity''

  10. Vibration characteristics of BWR primary containment vessel

    International Nuclear Information System (INIS)

    This paper reports on a part of a seismic proving test of a BWR Primary Containment Vessel (PCV), a vibration analysis that is carried out considering the effect of reinforcement around attached masses such as equipment hatches and a personnel airlock. A deflected shape with the local attached masses and reinforcement is expanded using free vibration modes of an axisymmetric PCV without attached masses and reinforcement. The free vibration modes are calculated considering a coupling effect between the PCV shell and suppression pool water using conical shell finite elements and ring fluid elements. The same expanding functions and free vibration modes are used to express strain energy, kinetic energy and work done by inertia forces due to input acceleration considering locally varying thickness of the PCV and attached masses. The equations of motion are obtained by substituting the strain energy, kinetic energy, and work done by inertia forces into the Lagrange's equation. Calculated natural frequencies, free vibration modes and frequency response functions are compared with numerical results obtained by a general shell finite element analysis and with test results using the large-scale high-performance vibration table of Tadotsu Engineering Laboratory, Nuclear Power Engineering Test Center (NUPEC). The comparison shows a reasonable agreement, leading to better understanding of the dynamic characteristics of the BWR PCV

  11. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

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

  12. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  13. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  14. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

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

  15. Un uso eficiente del agua doméstica : tratamiento de las aguas grises

    OpenAIRE

    Castro Montilla, María

    2015-01-01

    El trabajo expuesto a continuación plantea la reutilización de las aguas grises como un método novedoso para conseguir un uso eficiente del agua en el ámbito doméstico, llevando consigo un ahorro del recurso para conseguir, por una parte, la reducción de los costes debido a un consumo menor del agua; y por otra, una reducción de la huella ecológica para contribuir a un progreso hacia una arquitectura más sostenible y solidaria con el medioambiente. Se desarrollarán conceptos que abarcan, d...

  16. Repercusiones sanitarias de la calidad del agua: los residuos de medicamentos en el agua

    OpenAIRE

    Cristina Postigo Rebollo; Meritxell Gros Calvo; María José López de Alda Villaizán; Mira Petrovic; Antoni Ginebreda Martí; Damià Barceló Culleres

    2011-01-01

    Este trabajo resume varios estudios de monitorización de fármacos y drogas de abuso llevados a cabo en el medio ambiente acuático de las cuencas de los ríos Llobregat y Ebro con el fin de evaluar la calidad del agua en relación a la presencia de estas substancias y su potencial riesgo para la salud ambiental y pública. La identificación de estos compuestos aguas abajo del punto de vertido de las plantas depuradoras apunta a la descarga de agua residual tratada como la principal fuente de emis...

  17. Agua y tiempo en el arte

    OpenAIRE

    Pando Despierto, Juan

    1993-01-01

    Elemento básico de la división simbólica del mundo antiguo, el agua sigue siéndolo en la modernidad, y con una autoridad indiscutible, a la que su propia potencia —y su rareza esencial tratándose de agua dulce—, proporcionan un vigor estético y una fuerza geoestratégica planetarias. El arte hizo siempre del agua un factor de meditación y de representación, de concreción de los sueños del hombre. Su ductilidad plástica, propia para todos los marcos, todas las facturas, la hiz...

  18. Thermal aging evaluation of casting stainless steel under BWR environment

    International Nuclear Information System (INIS)

    Effect of thermal aging under BWR condition on material properties of casting stainless steel were evaluated by such as Charpy impact test, using replaced BWR component material. Solution heat treatment was performed to the same material and the material properties were obtained. Comparing each material test results, impact value of thermal aging material was lower than solution heat treatment material. By the results, thermal aging effect on material properties under BWR condition was confirmed. The material properties were compared with model equation using PLM evaluation and conservativeness of model equation was confirmed. (author)

  19. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

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

  20. Tratamiento biológico de aguas residuales industriales

    OpenAIRE

    Lobo, Cintia Cecilia

    2014-01-01

    A nivel mundial la disponibilidad de recursos hídricos constituye un tema de gran preocupación. En la Tierra solo el 2.53% del total de agua corresponde a agua dulce y se estima que solo el 15% del total de aguas residuales de América Latina recibe tratamiento. En Argentina la principal fuente de recursos hídricos son las aguas superficiales que representan un 80% del total de disponibilidad de agua; dentro de ellas se destaca la Cuenca del Plata que representa un 34% de la superficie tot...

  1. El riego con aguas de mala calidad en la agricultura urbana. Aspectos a considerar. II. Aguas residuales urbanas

    OpenAIRE

    Fernando Ortega Sastriques; Rosa Orellana Gallego

    2007-01-01

    El precio del agua potable es elevado y en muchas ciudades su disponibilidad está limitada y disminuirá en el futuro. La agricultura urbana es una concurrente en el uso del agua, en el futuro es muy probable que haya fuertes limitaciones para el riego con agua potable. Así es que para la producción de vegetales en medio urbano será necesario el empleo de agua de fuentes alternativas y no de los acueductos. Por lo general, el agua del subsuelo en el entorno de las ciudades, así como l...

  2. El riego con aguas de mala calidad en la agricultura urbana. Aspectos a considerar. I. Aguas salinas o alcalinas

    OpenAIRE

    Fernando Ortega Sastriques; Rosa Orellana Gallego

    2007-01-01

    El precio del agua potable es elevado y en muchas ciudades su disponibilidad está limitada y disminuirá en el futuro. La agricultura urbana es una concurrente en el uso del agua. En el futuro es muy probable que haya fuertes limitaciones para el riego con agua potable. Así es que para la producción de vegetales en medio urbano será necesario el empleo de agua de fuentes alternativas y no de los acueductos. La extracción de agua del manto freático para el riego es una práctica común q...

  3. Small scale BWR core debris eutectics formation and melting experiment

    International Nuclear Information System (INIS)

    A small scale experiment has recently been performed at Oak Ridge under the auspices of the BWR Severe Accident Technology (BWRSAT) program to provide information concerning the formation of mixtures during heatup of representative BWR reactor vessel bottom head debris and to determine the composition and melting temperatures of these mixtures. The initial structure of the bottom head debris layers modeled in the experiment was taken from the results of recent BWR Accident Response (BWRSAR) code predictions for the short-term station blackout accident sequence. The experimental results provide useful information concerning the mixtures formed and their proportions and properties. The observed run-off of a stainless steel-zirconium eutectic alloy supports the contention that the initial pour from a BWR reactor vessel would consists of molten metals at relatively low temperatures. (orig.)

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

    International Nuclear Information System (INIS)

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

  5. Trihalometanos en aguas de consumo humano

    Directory of Open Access Journals (Sweden)

    C Hernández Sánchez

    2011-01-01

    Full Text Available Los trihalometanos (THMs son subproductos de desinfección que resultan de la cloración de las aguas. Los THMs engloban a sustancias como el cloroformo (CHCl , el 3 bromodiclorometano (CHBrCl , el clorodibromometano (CHBr Cl 2 2 y el tribromometano (CHBr . La Agencia Internacional para la 3 Investigación del Cáncer ha clasificado al cloroformo y al bromodiclorometano en el grupo 2B que incluye a las sustancias posiblemente carcinogénicas en humanos. Debido a su toxicidad y a que el agua de consumo humano constituye la principal vía de exposición del hombre a estos compuestos, su presencia en las aguas de consumo humano está limitada por la Directiva Europea 98/83/CE a 100 ìg/L. En esta revisión se destacan los aspectos más relevantes sobre su toxicidad, reglamentación, determinación, concentración en aguas, variaciones estacionales y eliminación.

  6. Swedes repair BWR thermal fatigue cracks

    International Nuclear Information System (INIS)

    In connection with an accident at the Barsebaeck-2 NPP the causes of cracking in steel pipe fittings of the BWR type reactor cooling system are investigated. In the course of testing carried out by the methods of gamma radiography and liquid penetrant inspection the cracks 10-100 mm long and with depth up to 10 mm are found. The most of the cracks is concentrated in regions near pipe fittings in the direction of water stream flow. The cause of crack formation is the thermal stress arising during mixing the water with different temperatures in particular, the feedwater having at normal operational conditions temperature of 180 deg C and the emergency cooling system water with the temperature of 270 deg C. The conclusion is drawn on the necessity of designing the new configurations of joints which are able to withstand the temperature gradients

  7. Transportation activities for BWR fuels at NFI

    International Nuclear Information System (INIS)

    Nuclear Fuel Industries, LTD. (NFI) supplies fuel assemblies for both PWR and BWR in Japan. We can also manage transportation of the fuel assemblies from our fabrication facilities to the nuclear power plants of Japanese utilities. For the transportation of fuel assembly, we designed and fabricated the transportation containers to meet the requirements of the IAEA regulations, and licensed in Japan. This paper introduced the recent activity and R and D of NFI concerning transportation of BWR fuel assembly. NT-XII transportation container was developed for fresh BWR fuel assemblies. NT-XII container consists of inner container and outer container. Two BWR fuel assemblies with up to 5 wt.-% 235-U enriched are enclosed in an inner container. In the concept of NT-XII container design, we made the best priority to transportation efficiency, as well as ensuring fuel integrity during transportation. NT-XII has been used since 2002 in Japan. Thanks to the lightening weight of containers, the number of containers to be loaded to one transportation truck was increased up to 9 containers (equivalent to 18 fuel assemblies) compared with former type container (NT-IV transportation container) which can be loaded up to 6 containers (equivalent to 12 fuel assemblies). In addition to the design of brand-new container, we promote the improvement of the packaging methods. In Japan, in order to reduce the damage to the fuel rod and fuel spacer while transporting, polyethylene sleeves which are called 'packing separators' are inserted in the rod-to-rod gap of fuel assembly. However, packing separators requires time and cost for the installation at fuel fabrication facility. In the same way, huge time and cost are needed for removal of packing separators at nuclear power plant. For the improvement of preparation efficiency before and after transportation, we investigated the influence of vibration to fuel integrity in case of transportation without packing separators. Based on the above

  8. Radiation source term reduction in BWR plants

    International Nuclear Information System (INIS)

    This series of slides presents: the collective radiation exposures at US and European BWRs; the European experience with source term reduction measures (normal water chemistry - NWC): zinc addition, stellite replacement, full system decontamination; the effects of evolving water chemistries/US experience. The conclusions are summarized as follows: worldwide reduction of collective radiation exposures at BWRs by following the ALARA principle; zinc addition proven option for source term reduction for NWC and hydrogen water chemistry (HWC) plants; reducing feedwater iron has been proven to reduce dose rates - as operational observations in the US indicate; optimized feedwater iron is very important for fuel performance under all modes of water chemistry (HWC, Zn, and noble metal chemical addition (NMCA)); minimize 59Co sources/stellite, follow the ALARA principle; full system decontamination (FSD) plus zinc injection is an attractive option for reducing reactor coolant system (RCS) dose rates of mature BWR plants

  9. Improvement of BWR radioactive waste disposal system

    International Nuclear Information System (INIS)

    Description is made here about the improvement of the disposal systems for liquid and solid radioactive waste in recent BWR plants. Regarding the improvement of liquid waste disposal systems in Toshiba, emphasis was laid on crud removal and laundry drain treatment; as a result, a crud removal system utilizing a centrifugal separation mechanism and an evaporative enriching system using anti-foaming agent have been practicalized. An important problem left for solution is the method of volume reduction of solid waste. A method attracting attention of late is the plastic solidification which is far superior to the conventional cement and bitumen solidification. In Toshiba, this method is promising to be practicalized, in which the waste is dried and then solidified with thermosetting resin. (author)

  10. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  11. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  12. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  13. TRU transmutation type BWR fuel assembly

    International Nuclear Information System (INIS)

    The BWR fuel assembly is formed by bundling a plurality of fuel rods and a water channel disposed at the center of the assembly by a plurality of spacers. An upper tie plate and a lower tie plate are disposed to upper and lower portions of the fuel rods and the water channel respectively. An upper end plug of the water channel is attached detachably to a cylindrical main body of the water channel. A zircaloy tube incorporating TRU nuclides is contained and secured in the water channel. The zircaloy tube has such a structure as capable of incorporating and sealing oxides or metal materials containing TRU nuclides. Since the zircaloy tube containing TRU nuclides is contained not in fuel region but in the water rod, the loaded uranium amount of fuels is not reduced but the reactivity can be ensured. (I.N.)

  14. Contingency programs for BWR pipe cracking

    International Nuclear Information System (INIS)

    General Electric (GE) has aggressively addressed the problem of Intergranular Stress Corrosion Cracking (IGSCC). Intergranular Stress Corrosion Cracking has occurred in boiling water reactors in less than 1% of the Type-304 stainless steel welds in operating plants. However a comprehensive program by GE, with EPRI support in many cases, has provided technical solutions to limit IGSCC. As part of this program the Nuclear Services Department (NSD) has taken new technology and applied it to programs for the prevention of IGSCC and for repair of pipe cracks should they occur. The purpose of this paper is to provide a description and the current status of the major ongoing service programs for dealing with the problems related to IGSCC in BWR operating plants

  15. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical struct-ure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

  16. BWR startup and shutdown activity transport control

    International Nuclear Information System (INIS)

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 oF (

  17. In-reactor ECP measurements in BWR plants

    International Nuclear Information System (INIS)

    It has previously been confirmed that the initiation and propagation of Stress Corrosion Cracking (SCC) of stainless steel and Ni-based alloy exposed in the primary water of Boiling Water Reactor (BWR) depend on water chemistry. It is known that the corrosion environment is evaluated with Electrochemical Corrosion Potential (ECP) and SCC susceptibility is high when the ECP is high. Then it also has been confirmed that hydrogen injection in feedwater and Noble Metal Chemical Addition (NMCA) are effective to reduce ECP of reactor components. ECP measurements in BWR plants, which are BWR-3 and BWR-4, were performed in Normal Water Chemistry (NWC), Hydrogen Water Chemistry (HWC) and post-NMCA environments to evaluate the mitigation effect of SCC by HWC and NMCA. ECP measurements were conducted in the lower plenum region, bottom head region, below the core plate and near the bottom of the active fuel region, by installing modified LPRM with ECP electrodes, and at bottom head drain line flange location (BHDL). It is confirmed that the ECP are reduced to less than -200mV(SHE) by 0.9 ppm H2 concentration at feedwater in BWR-3 and 1.1 ppm H2 concentration at feedwater in BWR-4, and that the ECP can be reduced to less than -200mV(SHE) by 0.3 ppm or less H2 concentration after NMCA is applied. (author)

  18. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  19. BWR Source Term Generation and Evaluation

    International Nuclear Information System (INIS)

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-000061. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the Quality

  20. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  1. Sistemas de plantas de tratamiento de aguas residuales en Colombia

    OpenAIRE

    Lizarazo Becerra, Jenny Milena; Orjuela Gutiérrez, Martha Isabel

    2013-01-01

    El establecimiento de Sistemas de tratamiento de aguas residuales son importantes para proteger el Salud Pública y el medio ambiente. Si las aguas residuales van a ser vertidas a un cuerpo receptor natural (mar, ríos, lagos), será necesario realizar un tratamiento para evitar enfermedades causadas por bacterias y virus en las personas que entran en contacto con esas aguas, y también para proteger la fauna y flora presentes en el cuerpo receptor natural. El reúso del agua tratada, riego de áre...

  2. Arquitectura sostenible y conservación de agua

    OpenAIRE

    Lozano Portillo, Ana

    2008-01-01

    Esta monografía presenta la transcripción de la conferencia "Arquitectura sostenible y conservación de agua", pronunciada por Ana Lozano Portillo en el marco de la III Jornada Internacional sobre ahorro de agua organizada por el Ente Público del Agua de la Región de Murcia, celebrada el 12 de junio de 2008. Lozano Portillo, A. (2008). Arquitectura sostenible y conservación de agua. Cátedra Universidad Empresa Arquitectura Sostenible Bancaja Habitat. http://hdl.handle.net/10251/17962.

  3. Agua y emprendimiento social = Atrapanieblas y huertos urbanos

    OpenAIRE

    Cruz Gutiérrez, Abel

    2016-01-01

    El proyecto Acceso al Agua por Sistemas de Atrapanieblas de la Asociación “Peruanos sin Agua” con aguas no convencionales, Lluvias y Neblina. Por: Abel Cruz, Emprendedor Social y Líder del Movimiento Peruanos sin Agua El sistema atrapanieblas consiste en atrapar las gotas de agua microscópicas de la neblina a través del uso de mallas parecidas a una red. Las pequeñas gotas de la niebla quedan adheridas en los hilos de las mallas para luego ser redirigidas y acumuladas en depósitos. A...

  4. BWR operator training for emergency conditions using simulator in Japan

    International Nuclear Information System (INIS)

    BWR Operator Training Center Corporation (BTC) is the only training organization for all Japanese BWR utilities. BWR operator training program is constructed to maintain and upgrade the ability of the individual operator and the operation crews. The training course for the individual operator is focused on the operator's positions, from that of the main operator to the licensed shift supervisor. The target of the training for operation crews, which is called the family training, is to upgrade the team performance including shift supervisor. The training programs are combined to improve and upgrade the operator performance using simulator. BWR training simulator has been designed to demonstrate many kinds of malfunctions, modeling a single failure and multiple failure, including loss of coolant accident. It has also the alteration and scheduling function to simulate the abnormal and accident situation. These functions are utilized to easily simulate the accident scenario and enlarge the operator's knowledge in case of the abnormal condition. This paper describes the BWR operator training for emergency conditions using simulator and the advanced control board simulator performance including the abnormal situation functions. (author). 5 refs, 2 tabs

  5. BWR blowdown/emergency core cooling integral program

    International Nuclear Information System (INIS)

    The Program plan identifying the phased approach to testing has been completed and the first test phase is well underway. Test results to date show the expected lower peak cladding temperatures and slower system blowdown response for a BWR/6 LOCA simulation compared to the BWR/4. Counter-current flow limiting (CCFL) tests show that exact geometric replication was not necessary to simulate CCFL characteristics for a BWR prototype fuel bundle upper tie plate. Separate effects blowdown tests demonstrate the importance of flow length scaling in simulating break geometry for limiting the critical or blowdown flow rate. A new bundle thermal hydraulic method, MAYU04, has been completed. This method is shown to provide a substantial improvement in the prediction of bundle temperatures. The BD/ECC Program represents an important contribution to BWR safety research. Early results from the program have already provided a better understanding of the governing phenomena during hypothetical LOCA simulation tests. Future tests are expected to provide a basis for further improvements in BWR LOCA phenomena modeling

  6. Reliability of BWR high pressure core cooling

    International Nuclear Information System (INIS)

    The high pressure coolant injection system (HPCI), and the reactor core isolation cooling system (RCIC) are steam turbine driven systems that can inject water into a boiling water reactor at full operating pressure. Their purpose is to supply water during any failure that allows water to be lost while the reactor is at pressure and temperature. A large number of BWR plants are not meeting HPCI and RCIC performance goals for core cooling. NSAC considers concurrent failure of NPCI and RCIC to be the most probable potential cause of low reactor water level and possibly fuel damage in a boiling water reactor. Between January 1978 and May 1981, 169 licensee event reports were filed where HPCI or RCIC was inoperable or was declared inoperable. The present effort has shown that at least 40% of NPCI and RCIC problems might be averted by a high quality preventive maintenance program. About half of the plants do not perform cold quick-start surveillance testing of HPCI and RCIC. They do perform routine startup tests, but the equipment is first preheated and the startup is relatively gentle. However, emergency start-ups are abrupt and from the cold condition. Therefore, cold quick-start testing is the only way to assure that all components, control systems, and instruments are functioning correctly for automatic safety initiation. (author)

  7. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    An engineering plant analyzer has been developed at BNL for realistically and accurately simulating transients and severe abnormal events in BWR power plants. Simulations are being carried out routinely with high fidelity, high simulation speed, at low cost and with unsurpassed user convenience. The BNL Plant Analyzer is the only operating facility which (a) simulates more than two orders-of-magnitude faster than the CDC-7600 mainframe computer, (b) is accessible and fully operational in on-line interactive mode, remotely from anywhere in the US, from Europe or the Far East (Korea), via widely available IBM-PC compatible personal computers, standard modems and telephone lines, (c) simulates both slow and rapid transients seven times faster than real-time speed in direct access, and four times faster in remote access modes, (d) achieves high simulation speed without compromising fidelity, and (e) is available to remote access users at the low cost of $160 per hour. The accomplishment of detailed and accurate simulations in complex power plants at high speed and low cost are due chiefly to two reasons. The first reason is the application of five distinct modeling principles [2] which are not employed in any other simulation code. The second, and even more important reason is the utilization of a special-purpose peripheral computer with its 13 task-specific parallel processors

  8. Control method for BWR type power plant

    International Nuclear Information System (INIS)

    The present invention provides a method of controlling a BWR type plant having internal pumps capable of sufficiently utilizing the performance of a whole volume turbine bypass plant to enable stable supply of electric power upon load interruption of power generator thereof. Namely, upon occurrence of load interruption of a power generator or turbine trip, a plurality of internal pumps are tripped simultaneously to abruptly reduce a reactor core flow rate by a predetermined value or more. In this case, a reactor core flow rate abruptly reduction scram signal is prevented. Alternatively, a plurality of internal pumps are tripped simultaneously to abruptly reduce the reactor core flow rate. In this case, a reactor core flow rate abrupt reduction scram set value is changed in order to inhibit the reactor core flow rate abrupt reduction scram signal. With such procedures, upon load interruption of power generator or upon trip of turbine, reactor core flow rate is abruptly reduced by trip of internal pumps for avoiding increase of neutron fluxes due to reactor pressure change. However, since reactor scram is avoided, the operation can be continued upon load interruption of power generator. As a result, performance of whole volume turbine bypass plant can be utilized sufficiently even upon occurrence of load interruption of power generator. (I.S.)

  9. Analysis for steam separators in BWR

    International Nuclear Information System (INIS)

    Steam seperators in BWR are thin walled pipes with special cyclons for seperating steam and water arranged in a hexagonal scheme and welded to the core cover head. The pipes are connected to each other in two planes by a gridwork of small upright sheets also arranged in a hexagonal scheme. At the level of these gridworks there are also shroud rings for holding long screws. The sheets of the gridwork run concentric to the tubes and are interrupted by the tubes. The interiour forces of the gridwork-members are transferred through the tubewalls by circumferential bending which causes extreme elastic ovalizing of the pipe corss-section near to the gridwork connection. The height of these members may be constant or broader at the pipe connection (fig. 1d) for reducing the bending stresses in the pipe walls. If this ovalizing is involved in the elastic behaviour of the gridwork it works like a plate with a Poisson ratio ν approx. equal to 1. Such structures are extremely stiff against equal stresses in perpendicular directions and extremely weak for opposite stresses in both directions. Instead of this interrupted grid work the upper one may be a continues one, arranged tangentially to the tupes. The elastic behaviour of this gridwork is quite different to that of the interrupted one. (orig./GL)

  10. BWR control rod design using tabu search

    International Nuclear Information System (INIS)

    An optimization system to get control rod patterns (CRP) has been generated. This system is based on the tabu search technique (TS) and the control cell core heuristic rules. The system uses the 3-D simulator code CM-PRESTO and it has as objective function to get a specific axial power profile while satisfying the operational and safety thermal limits. The CRP design system is tested on a fixed fuel loading pattern (LP) to yield a feasible CRP that removes the thermal margin and satisfies the power constraints. Its performance in facilitating a power operation for two different axial power profiles is also demonstrated. Our CRP system is combined with a previous LP optimization system also based on the TS to solve the combined LP-CRP optimization problem. Effectiveness of the combined system is shown, by analyzing an actual BWR operating cycle. The results presented clearly indicate the successful implementation of the combined LP-CRP system and it demonstrates its optimization features

  11. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly of an BWR type reactor of present invention, in which a plurality of fuel rods are arranged in a regular square lattice like configuration include vibration-filled fuel rods in which granular nuclear fuel materials and granular non-nuclear fuel materials having a smaller neutron absorbing cross sectional area are mixed and filled. With such a constitution, the content of the mixed and filled non-nuclear fuel materials in the vibration filled fuel rods is at least 20% by a volume ratio in average in fuel assemblies. In addition, a burnable poison is optionally added and mixed to the granular mixture of the nuclear fuel material and the diluting granules. With such a constitution, the manufacturing cost can be reduced, and the combustion rate of the nuclear fission materials is increased to improve reactor core characteristics, thereby enabling to obtain sufficient Pu loading amount per assembly, and fuel assemblies excellent in flexibility in design and economic property can be obtained. (T.M.)

  12. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly has a 9 x 9 square lattice arrangement having a water channel which occupies an area of 3 x 3 lattice pattern corresponding to 9 fuel rods. Fuel pellets comprise those of not more 7 kinds which have fission products at enrichment degrees different by a spun of not less than 10%. Fuel rods comprise from 4 to 12 first type fuel rods and remaining second type fuel rods. The first type fuel rod is loaded with fuel pellets of fissionable products having an enrichment degree axially different at the upper and the lower portions. The second type fuel rod is loaded with fuel pellets of fissionable products having the same enrichment degree in the vertical direction. With such a constitution, the enrichment degree of fissionable products of fuel pellets in the fuel assembly for a BWR type reactor having different reactor constitution and operation conditions can be used in common. Accordingly, the degree of freedom for the design of the distribution of the enrichment degree is increased. (I.N.)

  13. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  14. Operation method for BWR type reactor

    International Nuclear Information System (INIS)

    In a BWR type reactor, the number of fuels at low enrichment, among initially loaded fuels, is increased greater than that of fuels to be exchanged, and the number of fuels at low enrichment remained in a reactor core after fuel exchange is decreased to smaller than that of entire control rods. Further, the fuels at low enrichment are disposed to the inner side except for the outermost circumference in the reactor core after fuel exchange. Since fuels of high reactivity are disposed at the outermost circumference in a second cycle, leakage of neutrons is increased and effective breeding factor is decreased. However, since the number of brought over fuels at low enrichment is decreased and the number of fuels at high enrichment is increased, effective average reactor core enrichment degree is increased, to compensate the lowering thereof due to the increase of neutron leakage. Since dispersion effect for the distribution of the enrichment degree can be utilized as much as possible by greatly reducing the number and the enrichment degree of fuels at low enrichment for initially loaded fuels, irrespective of the average enrichment degree and the fueling pattern in a first cycle, a burnup degree upon take-out of initially loaded fuels at ow enrichment degree can be increased to maximum. (N.H.)

  15. An overview of the BWR ECCS strainer blockage issues

    Energy Technology Data Exchange (ETDEWEB)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-03-01

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

  16. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

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

  17. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

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

  18. Examination of dissimilar metal welds in BWR and PWR piping

    International Nuclear Information System (INIS)

    This paper addresses dissimilar metal weld examinations at PWRS. Surveys were conducted to document the dissimilar metal weld configurations at PWR plants and to update the information known about dissimilar metal weld configurations at BWR plants. The experiences which BWR utilities have had with dissimilar metal weld examinations are documented and include: correct identification of IGSCC, indications thought to be IGSCC but were actually fabrication flaws, and difficulties encountered with the examination of dissimilar metal welds after stress improvement. An experimental program was conducted which verified that the longitudinal wave procedures developed for BWRs are also applicable to PWR designs

  19. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary

  20. Corrosion potential monitoring and its simulation in BWR conditions

    International Nuclear Information System (INIS)

    A simulation algorithm of corrosion potentials for BWR plant materials has been demonstrated. Cathodic and anodic electrochemical kinetic equations have been derived by analyzing kinetic models involving water radiolysis products of oxygen, hydrogen peroxide, and hydrogen. Corrosion rates of type 304 stainless steel with respect to corrosion potentials are formulated by numerical analysis as well. An electrochemical mixed potential theorem is applied to compute corrosion potentials. Flow rate effects of coolants on corrosion potentials of plant structural materials are expressed as a function of diffusion layer thickness. A fundamental technique and a theory to simulate corrosion potentials have been developed. Corrosion potentials in BWR conditions can be simulated by these results

  1. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

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

  2. Impact of Plant Noise on BWR Stability Analyses

    International Nuclear Information System (INIS)

    A small amount of thermal-hydraulic noise is present in a boiling water reactor (BWR). The noise originates primarily from minor fluctuations in fluid flow and pressure distribution in the recirculation system of the BWR and manifests itself as a small fluctuation on the order of 1-2% for the average power range monitors (APRM) during normal operation. A larger noise level is observed for single-loop operation than for two-loop operation. This noise has an impact on the stability performance of the BWR. This is particularly the case when the noise contains a significant component at the resonant frequency for BWR instabilities, which is typically on the order of 0.5 Hz. For a pump trip event that can lead to instability, the noise will impact the growth rate of the reactor instability. The initial magnitude of the oscillations will be larger as the decay ratio increases above unity. For operation at low flow, such as for minimum pump speed or single loop operation where the decay ratio is larger, the impact of noise could lead to small oscillations at the resonant frequency for the APRM signals. The impact of noise on BWR instabilities is analyzed with the TRACG code. TRACG consists of a multi dimensional two-fluid thermal hydraulics model and the three-dimensional kinetics model consistent with the GE 3D core simulator, PANACEA. TRACG models the reactor primary system and has been extensively qualified against test data and BWR plant data. Thermal hydraulic instability test data, as well as data from BWR instability events and tests, have been used extensively in this qualification. This paper demonstrates the impact of noise on BWR stability response for events leading to instability, such as pump trip events, as well as operation at low core flow due to single loop operation. The impact is illustrated through sensitivity studies with the TRACG code and by comparison to plant data. The impact of reactor noise on the performance of the instability detection system

  3. Full system decontamination experience in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, N.; Sugai, K.; Katayouse, N.; Fujimori, A.; Iida, K.; Hayashi, K. [Tokyo Electric Power Company, Tokyo (Japan); Kanasaki, T.; Inami, I. [Toshiba Corporation, Yokohama (Japan); Strohmer, F. [Framatome ANP Gmbh, Eelangen (Germany)

    2002-07-01

    At the Fukushima Daiichi Nuclear Power Station unit 3, unit 2, unit 5 and unit 1 of Tokyo Electric Power Company (TEPCO), the replacement of the core shroud and internals has been conducted since 1997 in this order. The welded core internals in operating BWR plants were replaced to improve stress corrosion cracking (SCC) resistance. At present these units are operating smoothly. The developed technology concept is to restore those internals in open air inside the reactor pressure vessel (RPV). To reduce the radiation dose rate inside the RPV, not only a shielding method was applied to cut the radiation from the irradiated structures but also a chemical decontamination method was applied to dissolve the radioactive crud deposited on the surface by using chemical agents. The calculated decontamination factor (DF) at the RPV bottom reached 35-117. As result, the dose rate decreased to approximately 0.1 mSv/h under water. Before and after the installation of the in-vessel shielding, a mechanical cleaning was extensively applied inside the RPV to remove the residual crud as well as the swarf, chips from cutting. As a result, the dose rate at the RPV bottom decreased to ranging from 0.2 to 0.4 mSv/h in air. A working environment for human access, which was better than expected, was established inside the RPV, resulting in 70, 140, 50 and 70 man-Sv (estimated) saving respectively at unit 3 (1F-3), unit 2(1F-2), unit 5(1F-5) and unit 1(1F-1). All four full system decontamination (FSDs) contributed to the successful realization of the core shroud replacement project under the dry condition in RPV.

  4. BwrCrud: Development and validation of a new code for improved simulation of activity transport in BWR primary systems

    International Nuclear Information System (INIS)

    During the nineties, much reactor experience and significant new findings have been acquired. Thus, it became necessary to improve the previous code for computer model. Development of an improved code (named as BwrCrud) for modeling of activated corrosion products in BWR systems is progressing to model the influence of factors such as iron and zinc flow, HWC operation and fuel failures in a better way. The present version of the Code, BwrCrud 1.0 was tested in the Kashiwazaki Kariwa 5 (KK-5) plant and it was concluded that the cord can work satisfactorily and manage to model the KK-5 behavior appropriately in consideration of some uncertainties in input and verification data. Phase 2 of the project is to be completed in the beginning of 1999. (M.N.)

  5. BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-07-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  6. De la pileta a las aguas abiertas

    OpenAIRE

    Ferreyra Balestra, Juan Pablo

    2013-01-01

    Dentro de las prácticas acuáticas, encontramos distintas disciplinas que la abarcan. Entre ellas encontramos a la natación tradicional (con sus 4 técnicas de nado y sus competencias), el waterpolo, nado sincronizado, saltos ornamentales, aguas abiertas, etc. Sobre esta última mencionada será el propósito de mi trabajo. Esta disciplina consiste en nadar en lugares abiertos como el mar, ríos, lagos y lagunas en lugar de hacerlo en una pileta. Aunque seamos expertos y llevemos años nadando en...

  7. Pensando al agua como parte del paisaje

    OpenAIRE

    Lorena Salvatelli

    2009-01-01

    El objetivo de este trabajo es discutir la importancia de los estudios de arqueología terrestre y subacuática como parte de un mismo proceso de interpretación del pasado. Se entiende que el registro arqueológico puede hallarse actualmente en una variedad de ambientes tanto terrestres como subacuáticos a causa de varias razones, como por ejemplo las oscilaciones de los niveles de agua ocurridas entre el Pleistoceno Final y el Holoceno Medio. Por lo tanto, se propone que para su ...

  8. Gente, tierra y agua en la Amazonia

    OpenAIRE

    Buitrago Garavito, Ana Isabel; Jiménez, Eliana; Tobón, Marco Alejandro; Gendrau Acho, Sheila; Ciro Rodríguez, Estefanía; Ciro Rodríguez, Alejandra; Trujillo Osorio, Catalina; Currea Dereser , Alejandra; Torres Zambrano, Nestor Ned; Serrato Hurtado, Clemencia; Santiago R. Duque; Moreno Hurtado, Flavio; Peñuela, Maria Cristina; Lloyd, Jon; Patiño, Sandra

    2008-01-01

    Con el ánimo de contribuir a la comprensión social, histórica y económica de la gente que habita la región amazónica y al entendimiento del funcionamiento de sus ecosistemas surge el libro Gente, tierra y agua en la Amazonia. lmani mundo III. Este libro es el tercero de la serie Imani mundo, la cual expone trabajos de investigación en la región amazónica tanto en aspectos sociales como en aspectos ecológicos, realizados por el instituto Amazónico de Investigaciones IMANI de la Universidad Nac...

  9. OBTENCIÓN DE LA CURVA DE DEMANDA DE AGUA TURÍSTICA PARA ACAPULCO, GUERRERO

    OpenAIRE

    Miguel Angel Cruz Vicente

    2011-01-01

    El agua se aprovecha en diversos usos que se diferencian por ser consuntivos y no consuntivos. Los primeros impactan en la disponibilidad porque aprovechan el agua y sólo retornan una parte de ésta (consumen agua), los no consuntivos retornan la totalidad del agua aprovechada (usan agua). En la clasificación del uso no consuntivo se encuentra el de recreación y turismo que se refiere a las actividades de contacto directo con el agua; por consiguiente, la disponibilidad de agua es básica para ...

  10. Pool swell in a nuclear containment wetwell. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, T.

    1976-04-01

    A brief description is presented of scale model tests conducted to study LOCA induced wetwell pool swelling in the BWR Mk 1 containment pressure suppression system. The Mk 1 containment configuration is described together with the scale model design, the conduct of the tests, and the experimental results. (DG)

  11. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  12. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

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

  13. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  14. Experience and reliability of Framatome ANP's PWR and BWR fuel

    International Nuclear Information System (INIS)

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  15. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    The BWR control rods made by ABB use boron carbide (B4C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B4C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  16. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

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

  17. Age-related degradation of BWR control rod drives

    International Nuclear Information System (INIS)

    This paper reviews the major age-related degradation mechanisms for U. S. boiling water reactor (BWR) control rod drives (CRDs). Component aging caused by various types of stress corrosion cracking, fatigue, general corrosion, wear, and rubber degradation are discussed. (author)

  18. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

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

  19. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  20. Assessment of the Prony's method for BWR stability analysis

    International Nuclear Information System (INIS)

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

  1. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

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

  2. Nuclear transmutation characteristics of reduced moderation BWR (Thesis)

    International Nuclear Information System (INIS)

    In the present thesis, the nuclear transmutation characteristics of reduced moderation BWR, which decides the spent fuel characteristics and its safety in its nuclear fuel cycle, were investigated and compared with other types of reactors. The major conclusions were obtained as follows: The decay heat and radioactivity from FPs increases in fuel burn-up. However, they which normalized with burn-up are small for the reactor with low specific power and long operation period due to the decay during the long operation period. Breeder type of reduced moderation BWR shows low decay heat and radioactivity from FPs because of the long operation period approximately 3000 days which realized by the high conversion ratio. That also shows low decay heat and radioactivity from actinide nuclides due to the hard spectrum. MA recycling reactor of high conversion type of reduced moderation BWR was designed. The neptunium, which has large impact for environmental burden from the viewpoint of nuclide transport analysis, can be incinerated approximately 40% of loaded inventory which corresponds to 22 units of LWR per year. LLFP (99Tc, 129I, 135Cs) transmutation by breeder type of reduced moderation BWR was estimated. As a result, the support factor cannot be lower than unity for each LLFP nuclides. In other words, the reduced moderation BWR cannot reduce LLFP because the LLFP target cannot be loaded inner of the reactor core due to the small margin of core specification. It is expected that these results and the characteristics of other types of reactor shown in the present study benefit the discussion for various nuclear fuel cycle options. (author)

  3. BWR 90+ - Nuclear power plant for 21st century

    International Nuclear Information System (INIS)

    BWR 90+ is a boiling water reactor, based on the previous models BWR90 and BWR75, and on the operational experiences gained with six reactors of the previous generation. The development work started in 1994 in co-operation with Teollisuuden Voima Oy (TVO). At present all the boiling water reactor owners participate the cooperation. The objectives of the development were: (1) to develop a boiling water reactor of competitive price level and short construction time, and which meets the latest safety requirements, (2) to itemize the technologies improving the security and competitivity of present plants, and (3) to maintain the expertise of the personnel of the companies participating the development work, and improving the BWR- technology. High power output and short construction time reduce the power generation costs. Large amount of fuel assemblies leads to higher safety margins. Reduction of scram groups from 18 to 16 reduces the amount of components, the assembly space and costs. The reactor technical data is as follows: Thermal power output 4250 MWth; electric power output 1500 MWe, construction time 1500 days, costs 1500 pounds/kWe, no. of fuel assemblies 872, no. of scram groups 16, turbines 1, the capacity factor 90% and the duration of service outage 3 weeks. Specific features of BWR90+ are: short construction time and low costs, risk for connection between wet and dry spaces has been minimized, reactor core remains covered by water during loss-of-coolant accident caused by fuel replacement, Passive collection and cooling of core-melt inside the containment, the containment is not the first wall against the spreading of core-melt, steam explosions and core- concrete interactions have low probabilities, high gas- volume of wet-space reduces the pressure increase during a severe accident, filter-equipped gas removal system forms the final overpressure shield, the containment is cylindrical, and the plant is equipped with digital instrumentation and control

  4. The Agua Salud Project, Central Panama

    Science.gov (United States)

    Stallard, R. F.; Elsenbeer, H.; Ogden, F. L.; Hall, J. S.

    2007-12-01

    The Agua Salud Project utilizes the Panama Canal's central role in world commerce to focus global attention on the ecosystem services provided by tropical forests. It will be the largest field experiment of its kind in the tropics aimed at quantifying the environmental services (water, carbon, and biodiversity) provided by tropical forests. The Agua Salud Watershed is our principal field site. This watershed and the headwaters of several adjacent rivers include both protected mature forests and a wide variety of land uses that are typical of rural Panama. Experiments at the scale of entire catchments will permit complete water and carbon inventories and exchanges for different landscape uses. The following questions will be addressed: (1) How do landscape treatments and management approaches affect ecosystem services such as carbon storage, water quality and quantity, dry- season water supply, and biodiversity? (2) Can management techniques be designed to optimize forest production along with ecosystem services during reforestation? (3) Do different tree planting treatments and landscape management approaches influence groundwater storage, which is thought to be critical to maintaining dry-season flow, thus insuring the full operation of the Canal during periods of reduced rainfall and severe climatic events such as El Niño. In addition we anticipate expanding this project to address biodiversity, social, and economic values of these forests.

  5. Reciclado de las aguas de decapado

    Directory of Open Access Journals (Sweden)

    López, F. A.

    1998-05-01

    Full Text Available Pickling liquors from steel industry, listed as hazardous wastes in most industrializated countries, have been treated by means of several procedures in order to recover the iron and other metals and to obtain an environmental friendly effluent. Depending on the treatment employed, different valuable ferrous products have been obtained. Ammoniojarosite and zinc ferrite were precipitated from the previously oxidised ferrous solutions.

    En la mayoría de los países industrializados, se incluyen las aguas de decapado, tanto las procedentes de los aceros al carbono como de aceros inoxidables, dentro de las listas de residuos tóxicos y peligrosos. Se presentan en este trabajo los resultados obtenidos en las investigaciones realizadas en los últimos años en los que se ha estudiado el tratamiento y recuperación de aguas de decapado de aceros comunes e inoxidables. Su tratamiento se ha enfocado hacia la consecución de dos objetivos: la recuperación del contenido total de metales y la obtención de un efluente de menor impacto ambiental. Se presentan los procedimientos estudiados para la obtención de compuestos de hierro, amoniojarosita y ferritas de zinc, con cierto valor añadido.

  6. LECTURAS DEL AGUA (SÍMBOLOS, ECOCRÍTICA Y CULTURA DEL AGUA

    Directory of Open Access Journals (Sweden)

    Eloy Martos Núñez

    2012-08-01

    Full Text Available Existen diversos paradigmas sobre el estudio del agua que inciden en ella desde distintos puntos de vista. La gestión del agua es abordada desde las ciencias físicas y geológicas, y se soporta sobre tecnologías como la ingeniería hidráulica. En el aspecto histórico-social, se subraya la importancia de considerar el agua no como un recurso como parte esencial de una cosmovisión que vertebra el patrimonio cultural de una comunidad. Por ejemplo, en el estudio de los desastres (inundaciones, plagas, etc. se ha puesto el énfasis en los datos hidroclimáticos, las respuestas sociales o la vulnerabilidad social. Sin embargo, la ecocrítica ha subrayado el papel de las fabulaciones folklóricas y de los Imaginarios en general para aportar una comprensión holística del problema. Los discursos fabulados admonitorios sobre las aguas y sus riesgos subrayan la importancia de estos lugares de memoria y de reforzar las identidades a través de estas ecoficciones y rituales. Se convierten, pues, en tradiciones útiles porque en realidad hablan de la sostenibilidad, la ética y la necesidad de preservar la memoria y la identidad cultural de la comunidad, frente a la depredación de la Naturaleza. De este modo, el conocimiento de esta cultura del agua no sólo contribuye a la adquisición de las competencias básicas de la educación, y en particular, a la educación literaria y del patrimonio, sino que contribuye a la gobernanza del agua y a la gestión sostenible y responsable de los recursos hídricos. Como indica E. Morin, la lectura y la literatura son una escuela de complejidad, y estos mitos inciden siempre en la multicausalidad de las acciones humanas.http://dx.doi.org/10.14572/nuances.v22i23.1749

  7. 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

    de Estabilidad que brinda la oportunidad de contar con datos reales de planta. Ademas de que en el caso 9 del ciclo 14 de este Benchamark se presentaron oscilaciones fuera de fase, las cuales resultan de gran interes debido a que los sistemas de deteccion que registran el promedio de la potencia del nucleo no las detectan. Adicionalmente en este trabajo se obtiene de manera individual el modelo de los lazos de recirculacion como un modulo independiente, ya que este modelo pertenece a otro trabajo y trabaja acoplado con la vasija del reactor. El modelo de los lazos de recirculacion es capaz de modelar varios transitorios de interes, como se muestra en el apendice A de este trabajo, entre los que se encuentran el disparo de bombas de recirculacion o la transferencia a baja o alta velocidad de ellas. El alcance del trabajo desarrollado se centra solamente en el nucleo del reactor, pero aprovechando la modularidad que ofrece PVM, es posible agregar componentes como separadores y secadores de vapor, lineas de agua de alimentacion y de vapor para obtener un modelo de un circuito cerrado completo. Las aplicaciones se centran principalmente en el entrenamiento de personal en la fenomenologia del BWR, y como herramienta de investigacion en el estudio de la dinamica de reactores BWR. El estudio de oscilaciones fuera de fase presenta actualmente retos como son la explicacion de la variacion de la linea neutra con el tiempo, oscilaciones fuera de fase no azimutales, sino axiales, etc. Por lo pronto se tiene un primer modelo orientado en esta direccion.

  8. El acceso al agua para consumo humano en Colombia

    OpenAIRE

    Martha Yánez Contreras; Karina Acevedo González

    2013-01-01

    El acceso al agua es un derecho humano que se debe garantizar sin ninguna discriminación. Este artículo examina la existencia de inequidades en el ejercicio de este derecho mediante un análisis discriminante que identifica los factores que explican las diferencias en la calidad del agua para consumo humano entre los municipios colombianos en 2009. Los resultados indican que existen inequidades en el acceso al agua potable. En particular, los municipios más pobres registran los niveles de cali...

  9. BWR-GALE, Radioactive Gaseous and Liquid Waste Release from BWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  10. Agua subterránea, deterioro de calidad y reserva

    OpenAIRE

    Auge, Miguel

    2008-01-01

    El agua en general y la subterránea en particular es un recurso limitado en nuestro planeta, donde más del 97% es salada y se concentra en mares y océanos. Del resto, alrededor del 2% constituye los casquetes polares en forma de hielo. Le sigue en magnitud el agua subterránea, cuya reserva hasta 1.000 m de profundidad se estima en el 0,5% del total, mientras que el volumen instantáneo de agua superficial llega a sólo el 0,02% y entre el agua del suelo y la atmosférica, componen algo más del 0...

  11. El agua, el TLC y el agro

    Directory of Open Access Journals (Sweden)

    Alfredo Suárez Mejía

    2006-01-01

    Full Text Available Este artículo de absoluta actualidad se re- fiere al tratamiento que el poder ejecutivo viene dando al tema del agro y los recursos de los cuales depende. Ya en prensa el tema de tierras, corrupción y violencia ha sido registrado en relación con el gobierno del presidente Uribe y el Incoder. Ahora se trata de examinar de qué manera el Incoder desmonta el manejo y propiedad de los distritos de adecuación de tierras de los cuales depende la actividad agropecuaria sin que para tal expropiación del recurso agua se requiera acudir a las instancias de un TLC. La hipertrofia de las posiciones contra el TLC ha permitido sin lugar a dudas que desde el derecho interno mientras se negocia se adelanten las normas que permiten las inversiones que interesan al espacio de la “integración comercial”

  12. Lahar hazards at Agua volcano, Guatemala

    Science.gov (United States)

    Schilling, S.P.; Vallance, J.W.; Matías, O.; Howell, M.M.

    2001-01-01

    At 3760 m, Agua volcano towers more than 3500 m above the Pacific coastal plain to the south and 2000 m above the Guatemalan highlands to the north. The volcano is within 5 to 10 kilometers (km) of Antigua, Guatemala and several other large towns situated on its northern apron. These towns have a combined population of nearly 100,000. It is within about 20 km of Escuintla (population, ca. 100,000) to the south. Though the volcano has not been active in historical time, or about the last 500 years, it has the potential to produce debris flows (watery flows of mud, rock, and debris—also known as lahars when they occur on a volcano) that could inundate these nearby populated areas.

  13. Pensando al agua como parte del paisaje

    Directory of Open Access Journals (Sweden)

    Lorena Salvatelli

    2009-12-01

    Full Text Available El objetivo de este trabajo es discutir la importancia de los estudios de arqueología terrestre y subacuática como parte de un mismo proceso de interpretación del pasado. Se entiende que el registro arqueológico puede hallarse actualmente en una variedad de ambientes tanto terrestres como subacuáticos a causa de varias razones, como por ejemplo las oscilaciones de los niveles de agua ocurridas entre el Pleistoceno Final y el Holoceno Medio. Por lo tanto, se propone que para su recuperación debe utilizarse una metodología adaptada al medio ambiente en el cual se encuentre. En este sentido, es importante la interpretación integral del paisaje para acceder a la comprensión de la diversidad ambiental en la que se puede emplazar un sitio arqueológico, y en consecuencia poder “bucear” un registro integrado.

  14. New accomplishments of Ambiente & Agua scientific journal

    Directory of Open Access Journals (Sweden)

    Getulio Teixeira Batista

    2009-04-01

    Full Text Available Ambiente & Água – An Interdisciplinary Journal of Applied Science (Ambi-Agua was classified as B2 in the Interdisciplinary Area, the major thematic area of this journal, by CAPES classification system known as "Qualis". Also, the journal was registered in the Digital Object Identifier (DOI® System. Three other facts were accomplished: 1 all published articles since April, 2008 have the submitted and accepted dates stamped; 2 long-term preservation will be now secured in association with DOAJ - Lund University Libraries and the e-Depot of the National Library of the Netherlands (KB; and 3 we entered into an electronic licensing relationship with EBSCO Publishing to have full text of our journal in EBSCO Publishing's databases in the near future.

  15. El agua, el TLC y el agro

    Directory of Open Access Journals (Sweden)

    Claudia Sampedro Torres

    2006-11-01

    Full Text Available Este artículo de absoluta actualidad se refiere al tratamiento que el poder ejecutivo viene dando al tema del agro y los recursos de los cuales depende. Ya en prensa el tema de tierras, corrupción y violencia ha sido registrado en relación con el gobierno del presidente Uribe y el Incoder. Ahora se trata de examinar de qué manera el Incoder desmonta el manejo y propiedad de los distritos de adecuación de tierras de los cuales depende la actividad agropecuaria sin que para tal expropiación del recurso agua se requiera acudir a las instancias de un

  16. Repercusiones sanitarias de la calidad del agua: los residuos de medicamentos en el agua

    Directory of Open Access Journals (Sweden)

    Cristina Postigo Rebollo

    2011-12-01

    Full Text Available Este trabajo resume varios estudios de monitorización de fármacos y drogas de abuso llevados a cabo en el medio ambiente acuático de las cuencas de los ríos Llobregat y Ebro con el fin de evaluar la calidad del agua en relación a la presencia de estas substancias y su potencial riesgo para la salud ambiental y pública. La identificación de estos compuestos aguas abajo del punto de vertido de las plantas depuradoras apunta a la descarga de agua residual tratada como la principal fuente de emisión de estos contaminantes en el medio acuático. El perfil de contaminación por fármacos fue bastante similar en ambas cuencas. No obstante, los índices de riesgo (HQ calculados para los fármacos en diferentes niveles tróficos (algas, dáfnidos y peces indican que los compuestos que presentan un mayor riesgo ecotóxico en el Llobregat son el sulfametoxazol (antibiótico sulfamida para las algas, el gemfibrozil (regulador de lípidos para las algas y los peces, el ácido clofíbrico (regulador de lípidos y la eritromicina (antibiótico macrólido para los dáfnidos, y el ibuprofeno (analgésico anti-inflamatorio para todos los eslabones tróficos. En el Ebro, los compuestos más problemáticos son el sulfametoxazol para las algas, y la eritromicina, el ácido clofíbrico y la fluoxetina (antidepresivo para los dáfnidos.Los niveles de drogas de abuso (y sus metabolitos determinados en la cuenca del Ebro son aproximadamente uno y dos órdenes de magnitud más bajos respectivamente que los determinados en las aguas de salida y de entrada a las depuradoras. Sin embargo, debido a la falta de datos sobre su ecotoxicidad, no se han podido calcular índices de riesgo.La presencia de estos compuestos (fármacos y drogas en aguas superficiales y de bebida aun no está regulada y, por lo tanto, no constituyen parámetros de obligado control. Aún es imprescindible seguir investigando para poder evaluar su posible efecto en la salud humana.

  17. Importance of momentum effects in BWR reactor vessel modeling

    International Nuclear Information System (INIS)

    The most severe class of events (other than design-basis LOCA's) for boiling water reactors (BWR's) entails rapid pressurization of the reactor vessel (RV) due to rapid valve closures in the main steamlines. The severity of these events is caused by the fact that the pressurization of the core results in a rapid decrease in the core void fraction and, coupled to a strong negative void coefficient of reactivity, in a rapid power increase if the reactor protective system is slow in responding. The DYNODE-B program models the nuclear steam supply system of BWR's. Early versions only explicitly calculated the RV dome pressure. The latest version permits optional explicit calculation of the dome, core outlet, and core average pressures. The use of the new option has shown the influence of the two-phase momentum effects on the core pressure and flow transients resulting from pressure wave reflections at the liquid (incompressible) core inlet region

  18. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

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

  19. Diagnosis of nonlinear BWR oscillations using TRAC/BF1

    International Nuclear Information System (INIS)

    The nonlinear nature of boiling water reactor (BWR) stability has been demonstrated in both experimental tests and lumped parameter calculational models. Point kinetic reactivity feedback is nonlinear because of its functional dependence on fuel temperature and moderator density. The TRAC/BF1 model used in this analysis differs from a lumped parameter model in its spatial extent. The model, intended to be consistent with a BWR/4, was developed with four active fuel channel components representing one hot, two average, and one peripheral bundles. The vessel internals were modeled explicitly. These internals include lower and upper plena, separator/dryers, core shroud, and dryer skirt. The jet pump/recirculation system is modeled in an azimuthally symmetric fashion. The feedwater and steam line boundary conditions are based on time-dependent data representative of that observed during the LaSalle oscillation event

  20. Coupled BWR calculations with the numerical nuclear reactor software system

    International Nuclear Information System (INIS)

    The Numerical Nuclear Reactor (NNR) is a software suite for integrated high-fidelity reactor core simulations including neutronic and thermal-hydraulic feedback. Using solution modules with formulations to reflect the multi-dimensional nature of the system, NNR offers a comprehensive core modeling capability with pin-by-pin representation of fuel assemblies and coolant channels. Originally developed for pressurized water reactors, the NNR analysis capabilities have recently been extended for boiling water reactor (BWR) applications as part of EPRI Fuel Reliability Program. The neutronics methodology is extended to treat non-periodic structure of BWR fuel assemblies, and a new Eulerian two-phase CFD boiling heat transfer model has been integrated with the software system. This paper summarizes the experience with, and results of, the first-of-a-kind coupled calculations as demonstration of a fully-integrated, high-fidelity simulation capability for assessment of margin to crud-induced failure from fuel-duty perspective. (authors)

  1. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil is under development. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of a severe accidents. (author)

  2. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil has been developed. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of severe accidents. (authors)

  3. Study on fast spectrum BWR core for actinide recycle

    International Nuclear Information System (INIS)

    A study has been performed on neutronic and thermal characteristics of a fast spectrum BWR core with tight fuel lattice for an innovative fuel cycle system named BARS (BWR with an Advanced Recycle System) aiming at Pu multi-recycling and MA's (Minor Actinides) burning. The coolant void reactivity tends to shift to positive direction due to the fast neutron spectrum, so the neutron streaming channel and axial enrichment distribution was introduced in the core design to overcome the tendency. It was found that negative void reactivity is attained for whole burnup cycle by using Monte Carlo code. As for the thermal characteristics, the critical power measurement test was performed in order to get the database and the result was compared with the Arai's critical power correlation. As the result, Arai's correlation can predict our critical power test data considerably. Through our core design study and benchmark tests, we will make the conceptual design of BARS core successfully. (authors)

  4. Neutron noise analysis of BWR using time series analysis

    International Nuclear Information System (INIS)

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

  5. Corrosion of Inconel X750 in simulated BWR core environment

    International Nuclear Information System (INIS)

    The corrosion tests of Inconel X750 in the simulated core environment by using an irradiated test loop were performed. From the experiments, the followings were clarified; 1. corrosion rate was proportional to minus cubic root of the time and the value was evaluated to be 19±4 mdm after one year, 2. the value was markedly high compared to other reported experimental data, but it had a good coincidence with the value evaluated from activity balance data in actual BWR plants, 3. the most important chemical specie which affected the corrosion behavior was identified to be hydrogen peroxide, and 4. the rate of specimens pre-filmed in the high temperature atmosphere at 700deg C for 5 hrs was reduced to be one third or less than that of specimens pre-oxidized in steam at 390deg C for 13 hrs which was the same condition for actual BWR fuel springs. (author)

  6. Electrochemical potential measurements under simulated BWR water chemistry conditions

    International Nuclear Information System (INIS)

    Laboratory studies have been performed to investigate the stainless steel corrosion potential under simulated BWR coolant chemistry conditions. In addition to dissolved oxygen and hydrogen, test parameters also included chemical additives, metallic ions and hydrogen peroxide at various concentrations. The effect of water flow velocity was also investigated under various water chemistry conditions. The details of test results have been described elsewhere, and the highlights of the investigation are summarized in this paper. (J.P.N.)

  7. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

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

  8. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    International Nuclear Information System (INIS)

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise

  9. BWR radiation control: plant demonstration. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  10. Valuation of BWR stability operating in natural circulation conditions

    International Nuclear Information System (INIS)

    Nowadays, the design of reactors having appropriate stability margins, the adoption of operating procedures avoiding possible unstable regions and the development of mitigation strategies to cope with inadvertent instability occurrences have strongly limited safety concerns in this regard. However, despite the obvious need for plant-specific Probabilistic Safety Assessment (PSA), BWR (boiling water reactor) transients of general interest can be identified and characterized as for example, overpressurization events, Large Break Loss of Coolant Accidents (LBLOCAs), feedwater temperature decrease, increase of core flow, main circulation pump flow rate increase, control rod withdrawal and others. Simulations of these complex scenarios have been improved by the utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes. In this work, the RELAP5/MOD3.3 thermal-hydraulic system code and the PARCS/2.4 3D neutron kinetic code have been adopted to predict the Peach Bottom BWR stability during recirculation pumps trip while the reactor is operating in a special region of power and core flow map. In the recirculation pump trip event, the stopping of the recirculation pumps causes a sharp decrease in the core flow, which generates a considerable negative reactivity insertion that tends to reduce power and, consequently, the amount of steam generated. The BWR reactor stability has been valuated during natural circulation conditions after the pumps trip event. The time evolution of the power and the related thermal-hydraulic parameters were investigated to analyse the behavior of the reactor for this special operation condition. (author)

  11. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

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

  12. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  13. El régimen jurídico de las aguas para usos agrícolas: aguas subterráneas

    Directory of Open Access Journals (Sweden)

    Fernando Brebbia

    2015-01-01

    Full Text Available Como ha señalado el profesor Adolfo Gelsi Bidart la relación entre tierra yagua es alguno de los aspectos que el derecho agrario debe tener en cuenta y ello importa establecer si la regulación de las aguas compete al derecho agrario, constituye una disciplina autónoma, o bien si su estudio corresponde a diversas ramas del derecho cada una de las cuales debe ocuparse de examinar determinados aspectos, o mejor, de distintas formas de su uso y aprovechamientol (...Contenido: Derecho agrario y derecho de aguas. Algunos datos estadísticos acerca del agua. Los usos del agua. Aguas públicas y aguas privadas. La utilización del agua para la agricultura. El tratamiento legal de las aguas para la agricultura en los países del Mercosur: Argentina, Brasil, Paraguay, Uruguay. Evolución legislativa del régimen de las aguas subterráneas en el derecho agrario comparado. Modernas tendencias sobre el dominio de las aguas subterráneas. Las leyes de aguas de España e Italia

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  15. Analysis of a BWR MKI utilizing advanced BMI-2104 computational techniques to calculate source terms

    International Nuclear Information System (INIS)

    The effects of incorporating code modifications to more realistically represent accident phenomenology were examined with specific focus on the BWR MKI AE γ-sequence. Results indicate that the source terms are only a fraction of those reported in BMI-2104 and that core/concrete interaction is not a significant source term contributor for the BWR MKI containments. (author)

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

    International Nuclear Information System (INIS)

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

  17. Panama Canal Watershed Experiment- Agua Salud Project

    Science.gov (United States)

    Stallard, Robert F.; Ogden, Fred L.; Elsenbeer, Helmut; Hall, Jefferson S.

    2010-01-01

    The Agua Salud Project utilizes the Panama Canal’s (Canal) central role in world commerce to focus global attention on the ecosystem services provided by tropical forests. The Canal was one of the great engineering projects in the world. Completed in 1914, after almost a decade of concerted effort, its 80 km length greatly shortened the voyage between the Atlantic and Pacific Oceans. An entire class of ships, the Panamax, has been constructed to maximize the amount of cargo that can be carried in a Canal passage. In today’s parlance, the Canal is a “green” operation, powered largely by water (Table 1). The locks, three pairs on each end with a net lift of 27 meters, are gravity fed. For each ton of cargo that is transferred from ocean to ocean, about 13 tons of water (m3) are used. Lake Gatún forms much of the waterway in the Canal transect. Hydroelectricity is generated at the Gatún dam, whenever there is surplus water, and at Madden Dam (completed in 1936) when water is transferred from Lake Alhajuela to Lake Gatún. The Canal watershed is the source of drinking water for Panama City and Colon City, at either end of the Canal, and numerous towns in between.

  18. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)]. E-mail: gepe@xanum.uam.mx; Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragan 779, Col. Narvarte, Mexico, DF 03020 (Mexico); Vazquez-Rodriguez, Alejandro [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)

    2006-09-15

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

  19. El grupo de trabajo del agua subterránea: experiencias de participación

    OpenAIRE

    Cisneros de Britto, Julio Cesar de; LÓpez Vera, Fernando

    2008-01-01

    López Vera, F., & Cisneros Britto, J. C. (2008). El Grupo de trabajo del agua subterránea: Experiencia de participación. Ciudadanos y usuarios en la Gestión del Agua. XIII Jornadas de Derecho de Aguas. Editor. Antonio Embid Irujo. Universidad de Zaragoza.

  20. Prevention of organic iodide formation in BWR`s

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  1. Prony's method application for BWR instabilities characterization

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  2. Prony's method application for BWR instabilities characterization

    International Nuclear Information System (INIS)

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

  3. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

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

  4. BWR/5 Pressure-Suppression Pool Response during an SBO

    OpenAIRE

    Javier Ortiz-Villafuerte; Andrés Rodríguez-Hernández; Enrique Araiza-Martínez; Luis Fuentes-Márquez; Jorge Viais-Juárez

    2013-01-01

    RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liqui...

  5. Characterization studies of BWR-4 neutron noise analysis spectra

    International Nuclear Information System (INIS)

    Neutron noise analysis measurements were made in three BWR-4 reactors under full-power conditions to determine the noise characterization spectra of the reactors with two different instrument-tube cooling configurations. Both configurations were designed to prevent flow-induced vibration of the instrument tubes and subsequent damage of fuel channel boxes caused by impacts of the tubes with the boxes. Noise spectra from these three reactors were compared with spectra previously obtained prior to changing the instrument-tube cooling configuration, and no evidence of impacting was found

  6. In-situ testing of BWR closure head studs

    International Nuclear Information System (INIS)

    Mechanized ultrasonic inspection of closure head studs often is on the critical path. In German BWR's, a floodcompensator is used which allows human access to the studs despite the water is up to a much higher level. For stud inspection this provides a potential solution to get out of the critical path. However, the space restrictions around the studs due to the geometry of the floodcompensator did not allow the use of the existing manipulators. This paper describes the design of a dedicated compact manipulator of a construction which copes with the restricted space available around the studs

  7. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  8. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly at the study on the effects of the radiation in the materials of the reactor; a little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear tracks manufactured in the ININ is presented, for the environmental monitoring in penetrations around the primary container of the Unit 1 of the Laguna Verde power plant. The monitoring of neutrons carried out with ends of radiological protection, during those operational tests of the reactor. (author)

  9. Neutron flux and fluence determination for BWR reactors

    International Nuclear Information System (INIS)

    Measurements of gamma emission rates from Fe and Cu dosimeters extracted from a BWR type reactor vessel were carried out in order to determine their total activity. The dosimeter's activity is related to the neutron flux there by taking into account the reactor material's embrittlement caused by neutron bombardment. The dosimeters were taken out after the first reactor operation cycle. From gamma radioactivity measurements of these dosimeters, neutron flux and fluence were calculated. These parameters are used in the determination of shift and adjusted reference temperature values needed for the development of pressure-temperature curves used during reactor operation

  10. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

  11. Corrosion products release from steel surface into BWR water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Korolev, A.S.; Berezina, I.G.; Sofyin, M.V.

    1986-02-01

    Factors influencing steel corrosion product release and transfer into a BWR primary circuit have been studied and reported on in this paper. The study of corrosion kinetics and corrosion product release was carried out on the samples tested under RBMK NPP condensate-feedwater cycle conditions, as well as, under test rig conditions. The ratio of corrosion product specific mass, transferred to the water, to the whole corrosion product specific mass of steel, formed under the given conditions was determined and used as a criterion, characterizing the extent of corrosion product transfer from the steel surface into the water.

  12. Development of underwater surface inspection system for BWR suppression chamber

    International Nuclear Information System (INIS)

    In order to inspect underwater paints surface of BWR suppression chamber as an in-service inspection, the underwater surface inspection system has been developed, which consists of magnetic crawler, crud recovery nozzles and special camera unit. The system can move on wall surface of pressure suppression chamber with magnetic crawler to inspect surface paints films under the muddy water with special camera unit after removing surface cruds. Based on inspection results, the paint films inspection criteria to assess degradation of paints films were proposed. (T. Tanaka)

  13. Los conflictos del agua en España

    Directory of Open Access Journals (Sweden)

    Gaspar Mairal Buil

    2005-04-01

    Full Text Available En este trabajo se analizan los conflictos producidos en España en los últimos años en torno a la construcción de grandes presas, tratando de mostrar cómo los afectados han forjado una noción compartida de riesgo y cómo ha surgido un movimiento social con un discurso proyectado hacia ámbitos más globales. Estos hechos se sitúan en relación con la política del agua de la nueva Directiva Marco del Agua (2000 de la Unión Europea para mostrar una insuficiencia, ya que ésta no toma en cuenta el significado cultural de los usos del agua.

  14. Sustancias húmicas en aguas para abastecimiento

    Directory of Open Access Journals (Sweden)

    Miller Camargo Valero

    2011-02-01

    Full Text Available Las sustancias húmicas hacen parte de los productos de degradación de la materia orgánica del suelo y se incorporan a las aguas superficiales por la acción de lavado que realizan las aguas de escorrentía. Estas sustancias han sido reconocidas como precursores en la formación de subproductos de la desinfección con cloro libre en plantas de tratamiento de agua potable. Los subproductos de desinfección, compuestos orgánicos halogenados, se han catalogado como sustancias potencialmente cancerígenas en humanos, y por tanto el interés en conocer más sobre las sustancias precursoras, mecanismos de formación de subproductos de desinfección, situación nacional y métodos para disminuir su formación.

  15. Defensa del agua desde la participación comunitaria

    Directory of Open Access Journals (Sweden)

    Martha Muñoz

    2013-09-01

    Full Text Available Las organizaciones comunitarias en Colombia defienden el mantenimiento de los páramos, indispensables para la provisión de agua en zonas rurales y urbanas. Se oponen a la idea del desarrollo económico que concibe al agua como servicio ambiental, vulnerando un derecho fundamental. Un ejemplo de organización es la que se lleva a cabo en el Santuario de Flora y Fauna de Iguaque. Ahí, los veedores saben que si no luchan por el agua perderán un derecho, como sucedió en las veredas de Olivos Alto y Bajo del municipio de Sanjuan de Rioseco, donde el Estado impuso sus políticas.

  16. Aguas subterráneas y humedales en zonas llanas

    OpenAIRE

    Custodio Gimena, Emilio

    2011-01-01

    Los humedales, en sus diversas formas, son importantes manifestaciones del agua en la Naturaleza. Proporcionan importantes servicios ambientales de gran interés para la vida y actividad humanas. Muchos humedales dependen en gran manera o totalmente del agua subterránea. Esto es tanto más cierto cuanto más árido en el territorio y entonces son ecológica y humanamente esenciales. Los humedales aparecen tanto en ambiente montañoso como de llanura. Aquí se da especial consideración a éstos últ...

  17. Tratamiento de aguas residuales textiles mediante un biorreactor de membrana

    OpenAIRE

    Lorena Salazar Gámez; Martí Crespi Rosell; Roberto Salazar

    2009-01-01

    En la actualidad, los investigadores dedicados al tratamiento de agua residual se han esforzado por obtener mejores rendimientos de depuración, reducción de costos de operación y mantenimiento en la generación de procesos que posean una elevada flexibilidad para soportar las variaciones en el afluente, que generen una mínima producción de lodos y que sean diseñados en condiciones mínimas de área. En este trabajo se expone una nueva tecnología en el tratamiento de agua residual, la de los Bior...

  18. IMPACTO DEL USO DE AGUA RESIDUAL EN LA AGRICULTURA

    Directory of Open Access Journals (Sweden)

    María E. Guadarrama-Brito

    2015-07-01

    Full Text Available La escasez de agua para uso urbano y la dependencia del riego para la producción agrícola en zonas de rápido crecimiento demográfico, han contribuido a que a nivel internacional aumente el interés en el reuso del agua. Existen riesgos para la salud debido a la presencia de microorganismos y contaminantes como los metales pesados y mutagénicos; los primeros impactan a corto plazo, debido a la contaminación de alimentos que pueden provocar, y los segundos impactan a largo plazo, contribuyendo a la salinización de suelos, lo que detrimenta la productividad para eventualmente derivar en el abandono de terrenos. El reuso del agua residual en la agricultura se ha convertido en una necesidad, la cual debe ser considerada como una alternativa, aunque no ha sido evaluada en los aspectos de contenido y migración de contaminantes, en particular de metales pesados. Algunos de los metales pesados pueden forman parte natural del suelo en cantidades que no resultan tóxicas para los seres vivos; sin embargo, la industrialización ha provocado un aumento de la presencia de estos en las aguas residuales que se utilizan para riego, con el consecuente riesgo para la salud humana y ambiental. El proceso de migración y fijación de contaminantes dentro de un sistema cerrado, dependerá de la capacidad de absorción por parte de los subsistemas agua-suelo-planta, aplicación de tasas de riego (concentración del contaminante, y de la persistencia y toxicidad de los contaminantes. El estudio aquí presentado, evalúa las tasas de migración de metales pesados presentes en un agua de riego, a través del sistema agua-suelo-planta; la evaluación se realizó con material del distrito de riego 028-Tulancingo que recibe aguas residuales de origen industrial, con presencia de Cobre, Manganeso y Zinc para diferentes grados de impacto.

  19. Sensor perfeccionado para la medida de la toxicidad del agua

    OpenAIRE

    Campo García, Francisco Javier del; Mas Gordi, Jordi

    2009-01-01

    La invención se basa en un sensor de medida para la toxidad del agua y consiste en una red de micro-electrodos recubiertos de una membrana y de un biofilm. La red de micro-electrodos puede ser estándar, y es posible su funcionamiento con un único electrodo, aunque ello acarrearía una perdida de sensibilidad. La membrana se usa para evitar la pasivación del electrodo por las bacterias presentes en el agua, y se utilizará en una de las realizaciones posibles "Nafion", aunque también es posible ...

  20. El estado actual de los ambientes del agua

    OpenAIRE

    Mateu Bellés, Joan F.; Camarasa Belmonte, Ana María

    2008-01-01

    Los ambientes del agua –recursos singulares por sus funciones en la naturaleza y por sus prestaciones y connotaciones para los grupos humanos- son un valioso patrimonio. En concreto, los lagos, los ríos y las masas marinas y subterráneas constituyen bienes territoriales que aportan diversidad ecológica, estructuran paisajes y, a menudo, son utilizados por los humanos para el desarrollo de sus actividades (Martín Montalvo, 1996). A lo largo del tiempo, los ambientes del agua no son estables e...

  1. Tratamiento de agua residual urbana con salinidad variable

    OpenAIRE

    Cort??s Lorenzo, Carmen

    2012-01-01

    Cuando los procesos biol??gicos se aplican al tratamiento de aguas residuales salinas, las altas concentraciones de sales inorg??nicas pueden afectar negativamente a la oxidaci??n de carbono, nitrificaci??n, desnitrificaci??n y eliminaci??n de f??sforo, e incluso pueden causar que los sistemas de tratamiento de aguas residuales no produzcan los rendimientos deseados. La sal puede ocasionar cambios en el metabolismo microbiano, en las propiedades de sedimentaci??n del lodo y de los fl??culo...

  2. Servicios ambientales, agua y economía

    OpenAIRE

    Juan Pablo Ruiz Soto

    2007-01-01

    Los servicios ambientales (SA) asociados a los ecosistemas naturales están seriamente amenazados por acciones locales y por el calentamiento global. El SA asociado al agua y su efecto sobre calidad y cantidad en cuencas hidrográficas aportantes a acueductos y sistemas de riego, son de gran importancia económica y social. Para Bogotá-Colombia el SA de los páramos y bosque nublado en la cuenca de Chingaza representa US$18,2 millones anuales. Para el abastecimiento de agua a la ciudad de Quito-E...

  3. Calidad del agua en marinas pequeñas

    OpenAIRE

    Fuentes Mariles, Óscar Arturo; Osnaya Romero, Javier; Magaña Melgoza, Pedro A.

    1997-01-01

    Se describe un método para calcular la concentración de un contaminante o bien la concentración de oxígeno disuelto (OD) y la demanda bioquímica de oxígeno (DBO) para una marina donde existe un mezclado completo de las sustancias en sus aguas. Los resultados del método permiten analizar la calidad del agua, revisar el tamaño de la marina, evaluar las consecuencias de la descarga de un contaminante, estimar la magnitud de bombeos para reducir la concentración de contaminantes, etc. También se ...

  4. Microorganismos indicadores de la calidad del agua potable en cuba

    OpenAIRE

    Marlen Robert Pullés

    2014-01-01

    Los agentes patógenos transmitidos por el agua constituyen un problema mundial que demanda un urgente control. Las bacterias, virus y parásitos causan enfermedades que varían en severidad. La determinación de microorganismos en el agua de consumo y su concentración proporcionan herramientas de control, indispensables para la toma de decisiones. Los controles rutinarios de todos los microorganismos, potencialmente riesgosos para la salud, resultan difíciles de llevar a cabo, debido a que ello ...

  5. Reuso de aguas residuales domésticas en agricultura. una revisión

    OpenAIRE

    Silva, Jorge; Torres, Patricia; Madera, Carlos

    2010-01-01

    Las aguas residuales son una importante fuente adicional para satisfacer la demanda del recurso, a causa de la disponibilidad limitada de agua potable para cubrir los requerimientos de las poblaciones, los bajos costos, los beneficios para los suelos agrícolas y la disminución del impacto sobre el ambiente. Sin embargo, el predominio del uso de aguas residuales crudas o diluidas con aguas superficiales y el bajo porcentaje de aguas residuales tratadas en Colombia y en los países de Amé...

  6. Determinación de la absorbancia Uv del agua sucusionada

    OpenAIRE

    Delgado Umaña, Mauricio

    2011-01-01

    Experimento donde se compara el efecto de las sucusiones sobre el agua midiendo la absorbancia Uv a 240, 300, 360 nm. Se encontraron diferencias estadísticamente significativas entre el agua sucusionada y el agua sin sucusionar. Los resultados contradicen los parámetros clásicos de la espectrofotometría al documentar absorbancias negativas. Se plantea que en el proceso de sucusionar el agua se establece una nueva fase del agua. / Abstract. Experiment which compares the effect of succussions o...

  7. Tratamiento de agua residual con elevado contenido de nitratos utilizando reactores biomembrana aireados

    OpenAIRE

    Lolmede, Ph.; Jácome, Alfredo; Vidart, T.; Tejero, I.

    2000-01-01

    Se describe la eliminación conjunta de carbono y nitrógeno de un agua residual, llevada a cabo en reactores biomembrana de laboratorio. El agua residual sintética tuvo una composición similar a la que resultaría de la mezcla, en determinadas proporciones, de agua residual urbana con agua de acuífero rica en nitratos. Se utilizaron dos reactores en paralelo, con igual superficie de soporte pero distinto volumen. Aunque se insufló aire a la biopelícula a través del soporte, el seno del agua ...

  8. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  9. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 1018 n/cm2) in the TRIGA Mark III Salazar reactor and separately with Ni+3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A2). (Author)

  10. BWR Servicing and Refueling Improvement Program: Phase I summary report

    International Nuclear Information System (INIS)

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

  11. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  12. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  13. BWROPT: A multi-cycle BWR fuel cycle optimization code

    International Nuclear Information System (INIS)

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes

  14. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  15. Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

    International Nuclear Information System (INIS)

    Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

  16. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

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

  17. BWR stability analysis with three-dimensional transient code

    International Nuclear Information System (INIS)

    Recently, neutron flux oscillations of two different modes were observed in several foreign BWR plants. One is core wide oscillation mode which is characterized by a phenomenon that neutron flux oscillates in-phase over a whole core. At La Salle 2 plant (U.S.A.), the amplitude of core wide neutron flux oscillation grew considerably large to result in a reactor scram, which aroused great concern about BWR stability. The other is regional oscillation mode which is characterized by the phenomenon, as typically observed at Caorso plant (Italy), that neutron flux of a half core oscillates out-of-phase to that of the other half core. These neutron flux oscillation phenomena were caused by nuclear-thermal hydraulic coupled instability and requires an evaluation study on oscillation detectability and effect on fuel integrity. Particularly, the regional oscillation mode requires three-dimensional analysis since it may bring about locally large amplitude power oscillation. For this reason, analysis was done with the three-dimensional transient code TOSDYN-2 to study reactor condition which causes the regional oscillation and also to evaluate fuel thermal margin under the neutron flux oscillations of these two instability modes. (author)

  18. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

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

  19. Analysis of NEACRP 3D BWR core transient benchmark

    International Nuclear Information System (INIS)

    NEACRP BWR cold water injection benchmark is analyzed by two codes: TRAC-BF1/SKETCH-N code system by JAERI, Japan and TRAB-3D code by VTT Energy, Finland. Basic features of the codes are described. Neutronics modules of the codes apply nodal methods; separate calculations are performed to compare their accuracy. Thermal-hydraulics modules are significantly different: TRAC-BF1 uses two-phase two-fluid model, while TRAB-3D applies drift-flux model with four separated equations. A representative set of the global and local reactor parameters is given for both the steady-state and transient conditions. TRAB-3D calculations have been performed with two slip correlations: EPRI and the simple Zuber-Findley correlation. A comparison of the two TRAB results shows the importance of the slip model on some computed reactor parameters. The results of the TRAC-BF1/SKETCH-N and TRAB-3D codes are in a close agreement, especially when the advanced EPRI correlation is used in the TRAB-3D code. The presented data can be useful for assessment of other BWR codes. (author)

  20. BWR Mark II ex-vessel corium interaction analyses

    International Nuclear Information System (INIS)

    This report describes the results of a series of studies conducted to investigate the behavior of core debris within a BWR Mark II containment. These studies focused on the interaction of core debris with concrete and steel structures (downcomers and inpedestal floor drains) within the drywell, the transport of debris through these drains and downcomers into the wetwell, and on debris-water reactions within the wetwell. Estimates of the conditions under which debris would penetrate the in-pedestal drain lines, the time-dependent behavior of the debris within the drain lines, and the amount of debris which might enter the suppression pool via these drain lines are provided. An assessment of the conditions under which the upper lip of the downcomers would be expected to fail (i.e. melt) due to exposure to hot core debris is presented. Finally, the unique characteristics of debris water interactions in Mark II containments are discussed, the existing knowledge base regarding core-concrete debris-water interactions is summarized, and an evaluation of the applicability of the MELCOR 1.80 code's debris-water interaction model to BWR Mark II's is presented

  1. TRAB, a transient analysis program for BWR. Part 1

    International Nuclear Information System (INIS)

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

  2. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

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

  3. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  4. Low power level safety management of Finnish BWR

    International Nuclear Information System (INIS)

    Good practices in work coordination and safety management have contributed to short refueling outage duration in Finnish BWR plants. Human and organizational factors are considered especially important in the low paper states, which consist of start-up, shut-down and the outage period itself. This originates from the use of external labour during the outage, the number of both contemporary and sequentially linked human actions and the variety of potential ways the personnel can affect the plant state. While the containment barrier does not exist, more organizational and administrative means have to be used in risk management. To promote the safety further, special studies have been carried out. This paper discusses both the low power mode PSA and the studies of work orientation and competence among the operating staff in Olkiluoto BWR plant. An advanced outage control requires also open-minded consideration of potential risks and the means for their reduction. Good results in low power risk management can be reached only by the involvement of both the plant operating and the maintenance staff. A profound safety management is a prerequisite for safe low power states. (author)

  5. BWR stability analysis with the BNL Engineering Plant Analyzer

    International Nuclear Information System (INIS)

    March 9, 1989 instability at the LaSalle-2 Power Plant and more than ninety related BWR transients have been simulated on the BNL Engineering Plant Analyzer (EPA). Power peaks were found to be potentially seventeen times greater than the rated power, flow reversal occurs momentarily during large power oscillations, the fuel centerline temperature oscillates between 1,030 and 2,090 K, while the cladding temperature oscillates between 560 and 570 K. The Suppression Pool reaches its specified temperature limit either never or in as little as 4.3 minutes, depending on operator actions and transient scenario. Thermohydraulic oscillations occur at low core coolant flow (both Recirculation Pumps tripped), with sharp axial or redial fission power peaking and with partial loss of feedwater preheating while the feedwater is flow kept high to maintain coolant inventory in the vessel. Effects from BOP system were shown to influence reactor stability strongly through dosed-loop resonance feedback. High feedwater flow and low temperature destabilize the reactor. Low feedwater flow restabilizes the reactor, because of steam condensation and feedwater preheating in the downcomer, which reduces effectively the destabilizing core inlet subcooling. The EPA has been found to be capable of analyzing BWR stability '' shown to be effective for scoping calculations and for supporting accident management

  6. BWR lower plenum debris bed models for MELCOR

    International Nuclear Information System (INIS)

    Work is underway at Oak Ridge National Laboratory (ORNL) to incorporate certain models of the Boiling Water Reactor Severe Accident Response (BWRSAR) code into a local version of MELCOR. Specifically, the BWR lower plenum debris bed and bottom head response models taken from BWRSAR are being tested within the local MELCOR code structure. Upon successful completion of testing, recommendations for formal adoption of these models will be made to the Nuclear Regulatory Commission (NRC) and to the MELCOR code development staff at Sandia National Laboratories (SNL). The SNL code development staff retain exclusive responsibility for maintaining the configuration control for the official version of MELCOR. The BWR lower plenum debris bed and bottom head response models permit the calculation of heatup, melting, and relocation of the debris after dryout. They predict the response of the lower plenum internal structures and the bottom head as well as the composition and timing of material release from the vessel. They have been previously applied in severe accident analyses for the Containment Performance Improvement (CPI) Program and the Mark I shell survivability study (NUREG/CR-5423), and in recent assessments of candidate accident management strategies. This paper provides a brief description of the purpose and operation of these models. 11 refs., 15 figs., 5 tabs

  7. Aging assessment of BWR control rod drive systems

    Energy Technology Data Exchange (ETDEWEB)

    Greene, R.H.

    1991-01-01

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

  8. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    A denatured (U-233/Th)O2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O2-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O2-fueled BWR should perform similar to a UO2-fueled BWR under all operating conditions. A (Pu/Th)O2-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO2-fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  10. Estudio de la calidad del agua de las fuentes de agua subterránea del Alto Palancia

    OpenAIRE

    RAMOS HERNÁNDEZ, RAQUEL

    2014-01-01

    En los últimos tiempos, la calidad del agua se está modificando debido a diferentes sustancias y acciones que la contaminan, y muchas de ellas son responsabilidad de la acción humana, principalmente a través de los vertidos incontrolados de la industria e incluso los municipales y procedentes de actividades como las agrícolas y ganaderas. Como objetivo prioritario para la salud pública está el control sanitario de agua para el consumo humano. Un enfoque integral de evaluación y...

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

    International Nuclear Information System (INIS)

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

  12. Water chemistry control practices and data of the European BWR fleet

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

  15. Dureza complexométrica del agua. Ejercicio 1.

    OpenAIRE

    Milla González, Miguel

    2013-01-01

    Se trata de un problema sencillo relacionado con la determinación complexométrica de la dureza del agua. Deberá calcularse los gramos de AEDT disódico necesarios para preparar un volumen dado del citado agente complejante de molaridad M. Los datos son aleatorios.

  16. Contaminantes en los cuerpos de agua del sur de Sonora

    Directory of Open Access Journals (Sweden)

    López-Ríos Olga

    2001-01-01

    Full Text Available Objetivo. Ahondar en el conocimiento sobre la incidencia de las actividades productivas del hombre en su medio, y el riesgo potencial que representa el deterioro ambiental para su propia salud y la sobrevivencia de especies animales. Material y métodos. Se utilizó el paquete informático Decision Support System for Industrial Pollution Control DSS (IPC, desarrollado por el Banco Mundial, la Organización Panamericana de la Salud (OPS y la Organización Mundial de la Salud (OMS, que se aplicó a un estudio de caso sobre los cuerpos de agua del sur de Sonora, durante 1996. Resultados. Estos sugieren que existen al menos tres factores que han contribuido a la desaparición de criaderos naturales de pescado y camarón en el municipio de Huatabampo: a la filtración de fertilizantes y pesticidas en las aguas de los drenes; b el asolvamiento producido por el depósito continuo, en el pasado, de agua de cola de la industria, c y la descarga inapropiada de aguas negras en los drenes. El texto completo en inglés de este artículo está disponible en: http://www.insp.mx/salud/index.html

  17. Necesidades de agua en explotaciones de vacuno lechero.

    OpenAIRE

    Callejo Ramos, Antonio

    2014-01-01

    Uno de los principales factores a considerar en el bienestar del ganado lechero es un adecuado acceso al agua como elemento imprescindible para la vida, además de ser necesario para el correcto estado sanitario de los animales y, por tanto, para lograr un adecuado nivel de producción.

  18. Uso Eficiente y Sustentable del agua. Implicancias y Condicionantes

    Directory of Open Access Journals (Sweden)

    Miguel Mathus Escorihuela

    2013-12-01

    Full Text Available El agua no es un recurso natural más, es un recurso vital que permite explicar integralmente los ciclos de la naturaleza y las interacciones entre las especies y el medio en que habitan en virtud del ciclo hidrológico. Pero el agua resiste al hombre, por lo que su aprovechamiento exige acciones concretas, que deben ser racionales y servir al bienestar de la sociedad, para poder alcanzar lo que hoy se llama calidad de vida en un ambiente sano. Para ello, el agua requiere de la Política Hídrica donde se determinan los fines; de la Planificación Hídrica racional y coherente que se encuentre volcada en Planes, Etapas y Programas; de la Legislación, que brinde el marco normativo, de la Administración que lleve a cabo la Política de la Gestión que es el ámbito neurálgico vinculado al uso eficiente del agua

  19. Radioactivity in drinking water Radiactividad en aguas de consumo

    Directory of Open Access Journals (Sweden)

    Enrique Estrada Vélez

    2011-12-01

    Full Text Available The European Directive on the quality of drinking water in 2003 establishing limits on the values of the radioactive parameters concerning the quality of drinking waters was translated to the Spanish legislation on 2003. These parameters are the gross alpha and beta indexes, tritium content and total indicative dose. The corresponding Spanish Royal Decree came into force in 2008. Not only tap drinking water is subjected to control of radioactivity but also commercial drinking water. Different estudies have been carried out all over the world in both cases. There is a wide range of techniques for the measurement of radioactivity in water. This article describes these techniques and shows when they must be applied. Finally the most important results obtained after analysis in tap water and commercial water are shown both in Spanish and non-Spanish waters. The results show that limits are overtaken in some cases and special care must be taken in order to reduce the levels of radioactivity in drinking water as much as possible.La directiva europea sobre calidad en el agua potable del año 2003 estableció los límites para los valores de los parámetros radiactivos. Esta directiva europea fue incorporada a la legislación 2003. Los parámetros a analizar son los índices alfa y beta total, el contenido en tritio y la dosis indicativa total. El correspondiente real decreto, en el caso de España, se comenzó a aplicar en el año 2008. No solamente el agua potable procedente de captaciones públicas está sujeta a control, sino que también se deben controlar las aguas comerciales. En ambos casos se han llevado a cabo estudios en todo el mundo. Existe un rango amplio de técnicas para la medida de la radiactividad en el agua de consumo. Este artículo describe tales técnicas e indica cuando se debe aplicar cada una de ellas. Por último, se muestran los resultados más importantes obtenidos tras el análisis de aguas potables y comerciales, tanto en

  20. Remoción de Metales de Aguas Residuales Industriales

    Directory of Open Access Journals (Sweden)

    Carlos Alberto Guerrero Fajardo

    2009-10-01

    Full Text Available En este trabajo se evalúa la tasa de remoción de cuatro metales pesados, comúnmente encontrado en Aguas Residuales Industriales (A,R,I,: Cobre, Cromo, Mercurio y Plata, mediante el empleo de plantas acuáticas vasculares "Jacintos de Agua". Las investigaciones para cada metal se realizaron, acondicionando aguas con cinco concentraciones diferentes que oscilaran por encima y por debajo del valor promedio encontrado en A,R,1,Se adoptó un modelo de comportamiento del sistema debido a la similitud que muestra la remoción de metales por el Jacinto de Agua, con el estudio de cultivos puros de MONOD, Esto permite realizar los ajustes estadísticos a los datosexperimentales encontrados en el presente proyecto. La ecuación establecida para el estudio de remoción de metales por un método biológico es el siguiente: % Remoción = A x t / B + t A= Porcentaje de remoción máxima B= Tiempo mínimo para alcanzar la mitad de la remoción máxima, t =Tiempo de remoción Las constantes A y B, indican la tendencia a la remoción y el tiempo mínimo de recolección, respectivamente. El Jacinto de Agua presenta limitación en su capacidad de remoción de metales, esto determina concentraciones máximas permisibles de metales en el agua para poder utilizar estas plantas acuáticas. El estudio de selectividad en la remoción de metales, indica que la presencia de Cromo, Mercurio y Plata disminuyen el porcentaje de remoción de Cobre, Se observa disminución en el porcentaje de remoción de Cromo, cuando están presentes Cobre, Mercurio y Plata en el agua residual. El caso del Mercurio es diferente a los dos anteriores, observándose más bien un efecto sinergético.

  1. Plant operation performance improvements of the General Electric (GE) boiling water reactors (BWR'S)

    International Nuclear Information System (INIS)

    This paper summarizes some of the plant operation performance improvement techniques developed by the General Electric Company Nuclear Energy Business Operation for the General Electric Boiling Water Reactors (GE BWR's). Through the use of both thermal and plant hardware operating margins, substantial additional flexibility in plant operation can be achieved resulting in significant improvements in plant capacity and availability factor and potential fuel cycle economics for the currently operating or requisition GE BWR plants. This list of techniques includes expanding the BWR thermal power/moderator flow operating domain to the maximum achievable region, operation with a single recirculation loop out of service and operation at rated thermal power with reduced feedwater temperatures. These plant improvements and operating techniques can potentially increase plant capacity factor by 1% to 2% and provide additional fuel cycle economics savings to the GE BWR's owners

  2. Rapid dewatering of Powdex resins at a BWR

    International Nuclear Information System (INIS)

    For most BWR's a large portion of their radioactive waste produced is water demineralization resins, both powdered and bead. In order to minimize the quantities of resins produced, proper demineralizer operation, volume reduction and minimization techniques are relevant to spent resin dewatering and packaging. To meet burial requirements spent resin needs to be dewatered and packaged properly. Methods of dewatering spent resins have included centrifuge separation and pulling water out of the resin with a diaphragm pump. Various vendors are offering systems that provide rapid dewatering and volume reduction using filter/liners in combination with vacuum pumps and air blowers. The various systems required a standardized test program for proper comparison and evaluation. The program is described in this paper

  3. Radial nodalization effects on BWR [boiling water reactor] stability calculations

    International Nuclear Information System (INIS)

    Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using ''representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs

  4. Cobra-TF simulation of BWR bundle dry out experiments

    International Nuclear Information System (INIS)

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

  5. Study of transient turbine shot without bypass in a BWR

    International Nuclear Information System (INIS)

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  6. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  7. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  8. Recent training technology of BWR operators using full scope simulators

    International Nuclear Information System (INIS)

    Nuclear power plants are being operated at high standards now in Japan. There are far few opportunities for operators to perform in challenging situations. To maintain skill and refinement, the simulator training is indispensable for them. BWR Operator Training Center (BTC) provides training courses according to the grade and duty of the operators. The training force constitutes of personnel from utilities', manufacturers' and also BTC-hired personnel. One of the big features of BTC training is composite team type. In this form of training, men from different plants make a team and help each other study. On the human factor viewpoint, error experience on simulators is one of the important items. Training on recognizing subtle symptom is an example of a recent development. Team training for actual crew is effective from various viewpoints. (author)

  9. Influence of local regulations on TN dual purpose BWR casks

    International Nuclear Information System (INIS)

    Transnucleaire (Paris, France) and Transnuclear, Inc. (Hawthorne, New York, United Sates) have both developed Pressurized Water Reactor (PWR) spent fuel casks for storage and transport purposes. The products are supplied in Europe by Transnucleaire and in the United States by Transnuclear, Inc. Now the TN Group is working on a design for Boiling Water Reactor (BWR) spent fuel assemblies: the TN 52 L cask is designed for transport and storage is Switzerland, the TN 68 cask is designed for transport and storage in the United States. For storage purpose, national regulatory requirements have to be met: each country has specific demands and criteria. As a consequence, differences between the TN 52 L design and the TN 68 design for rather similar contents appear in several fields: the design work, the licensing process, the manufacturing and the operational life. (author)

  10. Development of jet pump inspection equipments in BWR

    International Nuclear Information System (INIS)

    This paper describes development of the remotely operated equipments for jet pump ultrasonic testing (UT) in boiling water reactors (BWRs) to enhance the availability of operating nuclear power plants. Stress corrosion cracking (SCC) in the reactor internals has been a major concern in the BWR in recent years. The developed equipments can accomplish the appropriate positioning precision as an application of the Toshiba phased array immersion UT technique and enhance the jet pump inspection performance with a shorter duration and reducing the load for the installation of them. Three types of inspection equipments are developed to cover the outside and inside of the jet pump inlet mixer and the diffuser without disassembling the inlet mixer and the outside of the jet pump riser elbow. Their configurations and specifications are shown in the paper respectively. (author)

  11. TRAB - A transient analysis program for BWR. Part 2

    International Nuclear Information System (INIS)

    TRAB is a transient analysis code for BWRs developed at the Technical Research Centre of Finland. It models the phenomena in the interior of the BWR pressure vessel and in related subsystems. The core model of TRAB can be used separately for LWR modelling. For PWR modelling the core model of TRAB is connected to circuit model SMABRE to form the SMATRA code. This report is a user's manual and documents the structure, contents and preparation of input for TRAB. The structure of TRAB input is very flexible, featuring input groups and subgroups identified with keywords and given in any order as well as data items in free format, freely mixed with explanatory texts. Users interface of the code can be used for modelling within input: through normal input it is possible to create new submodels. These may be functional or tabulated dependencies of the code variables, different types of delays, or ordinary linear differential equations

  12. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

  13. Core concept for long operating cycle simplified BWR (LSBWR)

    International Nuclear Information System (INIS)

    An innovative core concept for a long operating cycle simplified BWR (LSBWR) is currently being developed under a Toshiba Corporation and Tokyo Institute of Technology joint study. In this core concept, the combination of enriched uranium oxide fuels and loose-pitched lattice is adopted for an easy application of natural circulation. A combination of enriched gadolinium and 0.7-times sized small bundle with peripheral-positioned gadolinium rod is also adopted as a key design concept for 15-year cycle operation. Based on three-dimensional nuclear and thermal hydraulic calculation, a nuclear design for fuel bundle has been determined. Core performance has been evaluated based on this bundle design and shows that thermal performance and reactivity characteristics meet core design criteria. Additionally, a control rod operation plan for an extension of control rod life has been successfully determined. (author)

  14. Dynamic safety systems in BWR plant safety systems

    International Nuclear Information System (INIS)

    Dynamic Safety Systems (DSSs) are reactor safety function systems that are functionally controlled using dynamic rather than static processes. All components including software, whose failure could result in a critical safety system failure, are operationally verified by hard-wired components. Dynamic Safety Systems have been enveloped in the United Kingdom by AEA Technology for use in gas cooled reactors. One such system, known as ISAT trademark, is described in this paper. Through use of scenario testing of a DDS emulator on a Boiling Water Reactor plant training simulator described in this paper. Through use of scenario testing of a DSS emulator on a Boiling Water Reactor plant training simulator, it is shown that a DSS can provide a cost effective safety system in BWR power plants

  15. Control method for water quality of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a method of suppressing radiation exposure upon periodical inspection of a BWR type reactor, suppressing leaching of radioactive materials deposited and activated on fuels, and reducing radioactive deposition on pipelines and equipments made of a carbon steel and austenite stainless steel. Namely, control of water quality described below is conducted under the conditions that the Ni metal ion concentration is from 2 to 10ppb and the Zn metal ion concentration of from 3 to 15ppb in reactor water. (1) controlling the water quality based on neutral/purified water during normal operation and upon injection of hydrogen, (2) using fuels having spring members made of a Ni based alloy processed by aging hardening in atmospheric air, (3) using reactor water recycling pipelines made of an electrolyzed and polished austenite stainless steel, and (4) using carbon steel or low alloy steel for pipelines and equipments of a reactor system. (I.S.)

  16. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10-3/demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  17. BWR Full Integral Simulation Test (FIST). Phase I test results

    International Nuclear Information System (INIS)

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report

  18. BWR Full Integral Simulation Test (FIST). Phase I test results

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W S; Alamgir, M; Sutherland, W A

    1984-09-01

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report.

  19. BWR plant analyzer development at BNL [Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    An engineering plant analyzer has been developed at BNL for realistically and accurately simulating transients and severe abnormal events in BWR power plants. Simulations are being carried out routinely with high fidelity, high simulation speed, at low cost and with unsurpassed user convenience. The BNL Plant Analyzer is the only operating facility which (a) simulates more than two orders-of-magnitude faster than the CDC-7600 mainframe computer, (b) is accessible and fully operational in on-line interactive mode, remotely from anywhere in the US, from Europe or the Far East (Korea), via widely available IBM-PC compatible personal computers, standard modems and telephone lines, (c) simulates both slow and rapid transients seven times faster than real-time in direct access, and four times faster in remote access modes, (d) achieves high simulation speed without compromising fidelity, and (e) is available to remote access users at the low cost of $160 per hour

  20. Obtention control bars patterns for a BWR using Tabo search

    International Nuclear Information System (INIS)

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

  1. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

  2. Spectral effects in cavitation of BWR jet pumps

    International Nuclear Information System (INIS)

    BWR jet-pumps vibrate violently when the recirculation drive flow becomes sufficiently high that cavitation occurs in the throat section at the local temperature of the suction flow. This process occurs rapidly as the fluid flows through the nozzle throat and while mixing of the drive and suction flows is in process. Continued mixing during transit through the throat eventually results in a pressure recovery to the saturation pressure and subsequent collapse of the voids near the entrance to the diffuser. The liquid surrounding a vapor volume, or void, performs work during void collapse. The result is that energy is deposited in the liquid and converted locally to an over-pressure, or sonic wave. The power contained in this wave may be very large, since it is deposited in the relatively short void collapse time. The frequency spectrum of the sonic excitation depends on the temporal nature of the collapse, including multiple events when the void is composed of a swarm of bubbles of various sizes. The frequency spectrum of a single spherical void collapse is calculated from the bubble collapse dynamics as it propagates in an increasing ambient pressure field that is a function of drive flow rate, Qd. It is shown that the void collapse is a tandem process in the time frame of the bubble swarm. These considerations lead to comparisons with strain-gauge data taken in BWR tests with the head off. Averaged micro-strain data is compared with the calculated values at several pump speeds, then fit to the data set. The normalized result is in excellent agreement, indicating that the void collapse mechanism is the most likely source of vibration excitation in the jet pump structure

  3. Spectral effects in cavitation of BWR jet pumps

    Energy Technology Data Exchange (ETDEWEB)

    Terhune, J.H.; Karim-Panahi, K. [GE Nuclear Energy, San Jose, CA (United States)

    1996-12-01

    BWR jet-pumps vibrate violently when the recirculation drive flow becomes sufficiently high that cavitation occurs in the throat section at the local temperature of the suction flow. This process occurs rapidly as the fluid flows through the nozzle throat and while mixing of the drive and suction flows is in process. Continued mixing during transit through the throat eventually results in a pressure recovery to the saturation pressure and subsequent collapse of the voids near the entrance to the diffuser. The liquid surrounding a vapor volume, or void, performs work during void collapse. The result is that energy is deposited in the liquid and converted locally to an over-pressure, or sonic wave. The power contained in this wave may be very large, since it is deposited in the relatively short void collapse time. The frequency spectrum of the sonic excitation depends on the temporal nature of the collapse, including multiple events when the void is composed of a swarm of bubbles of various sizes. The frequency spectrum of a single spherical void collapse is calculated from the bubble collapse dynamics as it propagates in an increasing ambient pressure field that is a function of drive flow rate, Q{sub d}. It is shown that the void collapse is a tandem process in the time frame of the bubble swarm. These considerations lead to comparisons with strain-gauge data taken in BWR tests with the head off. Averaged micro-strain data is compared with the calculated values at several pump speeds, then fit to the data set. The normalized result is in excellent agreement, indicating that the void collapse mechanism is the most likely source of vibration excitation in the jet pump structure.

  4. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m2. Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58Co and 60Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  5. BWR core shroud replacement. A cost-effective alternative

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking of stainless steel internals of BWRs is a longstanding and recognized problem. Well-engineered approaches have been developed by utilities and engineering contractors to address many of the known problems. One of these is cracking of the horizontal and vertical welds in the core shrouds. The approach being used world wide to address cracking in the horizontal welds is to install vertical tie rods and supplemental lateral restraints which provide redundant vertical and lateral support of the shroud and thereby structurally replace the function of the shroud horizontal welds. This approach has been thoroughly analyzed, approved by regulatory bodies and implemented in the United States, Europe, and Taiwan in over 16 BWR units. Importantly, many of these tie rod installations have been installed on a preemptive basis as a prudent preventive maintenance action. As BWR plants age, additional cracking of core shrouds, core supports, core spray piping and jet pumps has been experienced at some plants, indicating the need for a cost-effective, technically-acceptable approach for dealing with these problems. One such approach is complete replacement of the core shroud assembly, including core supports and attached core spray piping. This operation also makes the jet pumps accessible for replacement if this is determined to be necessary. A complete replacement by welding has been accomplished in Japan, and provides one acceptable replacement method. An alternative replacement concept, described in this paper, takes advantage of the proven tie rod modification to provide an approach which can be accomplished without in-reactor welding and without draining the reactor vessel. As described in this paper, this approach can be accomplished with significantly reduced radiation exposure to workers, reduced costs, and in a fraction of the time required for the fully-welded approach. An additional benefit is that as-welded joints in the reactor

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  7. ROSA-III experimental program for BWR LOCA/ECCS integral simulation tests

    International Nuclear Information System (INIS)

    This is the final report of the ROSA-III experimental program, in which the summary of integral simulation test results is described on thermal-hydraulic behavior during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) and on the effectiveness of the emergency core cooling system (ECCS). Also presented in the report are the assessment results of computer codes for the BWR LOCA analysis and of the similarity between ROSA-III test results and thermal-hydraulic phenomena during a BWR LOCA by using ROSA-III test data and code analysis results. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core consisting of four half-length bundles. Many test series were conducted between April 1978 and March 1983. The similarity between a ROSA-III test and a BWR LOCA concerning the fundamental thermal-hydraulic phenomena has been confirmed for major ROSA-III tests. The accident scenario has been well understood and defined for various break locations and break sizes. The effectiveness of the current BWR ECCS design has been well demonstrated. (author)

  8. Investigation of BWR [boiling water reactor] instability phenomena using RETRAN-03

    International Nuclear Information System (INIS)

    In 1988, LaSalle, a boiling water reactor (BWR)/5, experienced severe flux oscillations following a trip of both recirculation pumps. The flux oscillations were terminated by an automatic scram at 118% of rated neutron flux. As a result of this event, the U.S. Nuclear Regulatory Commission has asked the BWR utilities to develop procedural or hardware changes that will assure protection of all safety limits. The rapid growth of the oscillations at LaSalle, and the fact that previous stability analyses had predicted the plant to be very stable, emphasizes that a better understanding of this phenomenon is needed before the success of the long-term fixes can be assured. The intent of the Electric Power Research Institute's work was to use BWR transient methods to model reactor instabilities and investigate the factors that dominate this phenomenon. The one-dimensional transient code RETRAN-03 (Ref. 1) was used. The following conclusions are drawn: (1) RETRAN has demonstrated the ability to model BWR instability (nonlinear oscillations). (2) The general system behavior predicted by RETRAN in BWR stability analyses matches theoretical prediction and plant data. (3) These one-dimensional, time-domain results have increased the understanding of BWR stability phenomena and have helped optimize the long-term solutions being developed by the utilities

  9. Agua y abastecimiento: gestión de cuerpos de agua en la ciudad de San Luis Potosí (México, 1831-1887

    Directory of Open Access Journals (Sweden)

    Yuritzi Hernández Fuentes

    2015-01-01

    Full Text Available Este trabajo plantea una aproximación sobre la gestión de cuerpos de agua y de los sistemas hidráulicos en la ciudad de San Luis Potosí (México, durante el período 1831-1887, a partir de las medidas que tomó el gobierno municipal para atender las necesidades de abastecimiento de agua. El texto analiza dos proyectos importantes en el manejo del agua: el acueducto La Cañada del Lobo y las acciones sobre las derivaciones de aguas a través de La Corriente. Los dos proyectos enfrentaron diversos problemas, entre ellos, el brote de enfermedades asociadas a los cuerpos de agua y la escasez del flujo de líquido a través del acueducto de La Cañada del Lobo.

  10. Detección de mutacarcinógenos y genotoxicidad en aguas residuales que surten dos plantas de tratamiento de agua de consumo

    Directory of Open Access Journals (Sweden)

    Naranjo LC.

    2001-06-01

    Full Text Available Las aguas residuales están cargadas de gran variedad de mutacarcinógenos entre ellos Hidro-carburos Policíclicos aromáticos, nitrosaminas y aminas Heterocíclicas todos ellos formadosdurante la combustión de material orgánico y en alimentos proteicos cocidos a altas tempe-raturas. Dichos mutacarcinógenos llegan a las aguas residuales especialmente a través de laorina y heces. Las aguas residuales procedentes de la población del Retiro son vertidas al ríoPantanillo después de ser tratadas en una planta a base de lodos activados. Este río surte repre-sas que alimentan las plantas de tratamiento de agua de consumo la Ayurá y Rionegro quedistribuyen agua potable a más de 2’000.000 de habitantes.

  11. The possibility and the effects of a steam explosion in the BWR lower head on recriticality of a BWR core

    International Nuclear Information System (INIS)

    The report describes an analysis considering a BWR postulated severe accident scenario during which the late vessel automatic depressurization brings the water below the level of the bottom core plate. The subsequent lack of ECCS leads to core heat up during which the control rods melt and the melt deposits on the core plate. At that point of time in the scenario, the core fuel bundles are still intact and the Zircaloy clad oxidation is about to start. The objective of the study is to provide the conditions of reflood into the hot core due to the level swell or a slug delivered from the lower head as the control rod melt drops into the water. These conditions are employed in the neutronic analysis with the RECRIT code to determine if the core recriticality may be achieved. (au)

  12. Relación entre aguas subterráneas y humedales

    OpenAIRE

    Custodio Gimena, Emilio

    2011-01-01

    Numerosos humedales dependen parcial o totalmente de las aguas subterráneas, así como de la orla de vegetación que los rodea. Ello es tanto más cierto cuanto más árido es el clima. Incluso parte de los aportes más o menos permanentes de agua superficial son agua subterránea en origen. Por lo tanto, la existencia, fluctuación y calidad mineral del agua de esos humedales depende de esas aguas subterráneas, a veces de forma muy específica. La explotación del agua subterránea de los acuíferos ...

  13. Investigación sobre tratamientos de aguas seguros en emergencias

    OpenAIRE

    Ruiz García, José Luis

    2016-01-01

    Esta Tesis Doctoral tiene como principal objetivo el obtener una cadena de tratamientos seguros de aguas seriados que nos permita asegurar la calidad de las aguas para consumo humano en caso de emergencias, de tal forma que se minimicen los efectos de acciones hostiles, como sabotajes o actos terroristas, desastres naturales, etc y buscar soluciones adecuadas para garantizar en este caso la salud. Las plantas de tratamientos de aguas existentes comercialmente no aseguran dicha ...

  14. Tratamiento de aguas industriales mediante reactor biológico de membranas

    OpenAIRE

    Aznar Jiménez, Antonio

    2008-01-01

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

  15. Proyecto de abastecimiento de agua a Herrera de Ibio

    OpenAIRE

    Baldó García, Ignacio

    2015-01-01

    El Trabajo de Fin de Grado realizado es un proyecto del abastecimiento de agua a la población de Herrera de Ibio, municipio de Mazcuerras en la Comunidad Autónoma de Cantabria, España. Se desarrollan todas las partes que componen un abastecimiento de agua: • Captación • Red de tuberías • Depósito La captación se realiza de un manantial de montaña a una cota de a 277 m. Desde ese punto parte la tubería en presión que une la captación con el depósito y tiene una longitud de 1.048 metros. El inf...

  16. DEMANDA DE AGUA PARA USO RESIDENCIAL Y COMERCIAL

    OpenAIRE

    Sandra Gabriela Gómez-Ugalde; Ramón Valdivia-Alcalá; José Alberto García-Salazar; José Saturnino Mora-Flores

    2012-01-01

    La escasez de agua y la sobreexplotación de los mantos acuíferos es un problema de índole ambiental y económico que afecta a la población de Texcoco, Estado de México. Con el objetivo de cuantificar la demanda de agua en el sector doméstico y comercial y las variables que la determinan, ante la problemática de sobreexplotación de los mantos acuíferos de Texcoco, se plantearon dos modelos econométricos de regresión múltiple para determinar si una política de precios es adecuada para la adminis...

  17. Parámetros de control de los procesos de regeneración del agua

    OpenAIRE

    Gómez López, Marc

    2008-01-01

    El agua regenerada es aquella agua proveniente de un proceso de depuración a la cual se le aplica un tratamiento adicional (llamado de regeneración) que permite devolverle, parcial o totalmente, el nivel de calidad que tenía antes de ser utilizada. Para que estas aguas constituyan un recurso hídrico a tener en cuenta, es necesario realizar una reutilización planificada que consiste en establecer unas normas de calidad en función del uso final del agua regenerada así como un cor...

  18. Diseño, monitorización y control para un tanque de agua potable

    OpenAIRE

    Fernández Varo, José Antonio

    2005-01-01

    Este proyecto trata de resolver la problemática que existe en la planta de tratamiento de agua de la Central Térmica de Ciclos Combinados (CTCC) de San Adrián del Besós. Debido al poco cloro en el suministro de agua de red, existe la posibilidad de que se produzcan brotes de legionela en el tanque de agua potable. Por este motivo y por la normativa vigente (DOGC núm. 3652/02, RD 140/03), se obliga a tener un control sobre el agua almacenada. A partir de este hecho, se han realizado estudio...

  19. El agua virtual: una oportunidad para la Argentina en el contexto internacional actual

    OpenAIRE

    Jaime, Ana María

    2013-01-01

    En forma recurrente, se considera que el agua desempeñará en el siglo XXI, el rol que tuvo el petróleo en el siglo XX. La diferencia fundamental entre ambos recursos es que el agua es un “recurso finito, vulnerable y esencial para mantener la vida, el desarrollo y el medio ambiente”, según la definición de la Conferencia Internacional sobre el Agua y el Medio Ambiente realizada en la ciudad de Dublin en el año 1992. El agotamiento y la creciente escasez de agua en el mundo, el cambio climátic...

  20. Comentarios sobre el comercio y mercados del agua subterránea en Canarias

    OpenAIRE

    Custodio Gimena, Emilio

    2011-01-01

    Los mercados del agua subterránea son una interesante posibilidad para la gestión del agua subterránea. Estos mercados han existido desde hace muchas décadas en las islas principales del Archipiélago Canario y han contribuido en parte al importante proceso de desarrollo del último medio siglo. Con variantes de isla a isla y de un lugar a otro, los mercados de agua subterránea son la superposición de tres elementos, el de propiedad del agua a modo de participaciones y acciones que se compran y...

  1. Como funciona una estación depuradora de aguas residuales (EDAR)

    OpenAIRE

    2010-01-01

    El video presenta, en lineas generales, el funcionamiento de una depuradora de aguas residuales urbana convencional de una gran población. En él se describe el funcionamiento tanto de la linea de tratamiento de las aguas residuales como la de los lodos generados en el proceso. Este tipo de plantas implican la aplicación al agua residual de tratamientos tanto desde el punto de vista físico-químico como biológico para obtener un agua tratada que cumpla con la legislación vigente. Asimismo, el t...

  2. Desarrollo de catalizadores para la eliminación de nitratos en agua de consumo humano

    OpenAIRE

    Jaworski, María

    2012-01-01

    Es bien sabido que la contaminación de las aguas puede proceder de fuentes naturales o de actividades humanas. En la actualidad la más importante, sin duda, es la provocada por el hombre: el desarrollo y la industrialización supone un mayor uso de agua. Así, las actividades de agricultura y ganadería, producen vertidos de pesticidas, fertilizantes y restos orgánicos de animales y plantas que contaminan de una forma difusa, pero muy notable las aguas, en particular las subterráneas. Las aguas ...

  3. Subproductos halogenados de la cloración en el agua de consumo público

    OpenAIRE

    Calderón J.; Capell C.; Centrich F.; Artazcoz L.; González-Cabré M.; Villalbí J.R.

    2002-01-01

    Objetivo: La presencia de trihalometanos en el agua de diversas ciudades españolas está demostrada. Este estudio pretende analizar sus concentraciones en el agua de la ciudad de Barcelona. Métodos: Se analizan 151 muestras de agua recogidas durante el año 1998. Los análisis se realizaron en el laboratorio municipal mediante purge and trap y posterior determinación por cromatografía de gases acoplada a espectrofotometría de masas. Los resultados se presentan según el origen del agua: ríos Llob...

  4. Impacto de la producción lechera en la calidad de los cuerpos de agua

    OpenAIRE

    Rafael Arocena; Guillermo Chalar; Carlos Perdomo; Daniel Fabián; Juan Pablo Pacheco; Mauricio González; Vanesa Olivero; Macarena Silva; Verónica Etchebarne

    2013-01-01

    La producción lechera es de las actividades de mayor desarrollo y más contaminantes en Uruguay. A efectos de evaluar y mitigar su impacto en la calidad del agua, se relacionan el modo de producción y de manejo de los efluentes lecheros con diversos indicadores de calidad de los arroyos receptores, en diez microcuencas de un embalse reserva de agua potable para Montevideo. Se analizó el agua subterránea en pozos, el contenido de fósforo del suelo, la calidad del ambiente y del agua de los curs...

  5. Impacto de la producción lechera en la calidad de los cuerpos de agua

    OpenAIRE

    Arocena, Rafael; Chalar, Guillermo; Fabián, Daniel; Pacheco, Juan Pablo; González Piana, Mauricio; Olivero, Vanesa; Silva, Macarena; García, Patricia; Etchebarne, Verónica; Perdomo, Carlos H.

    2012-01-01

    La producción lechera es de las actividades de mayor desarrollo y más contaminantes en Uruguay. A efectos de evaluar y mitigar su impacto en la calidad del agua, se relacionan el modo de producción y de manejo de los efluentes lecheros con diversos indicadores de calidad de los arroyos receptores, en diez microcuencas de un embalse reserva de agua potable para Montevideo. Se analizó el agua subterránea en pozos, el contenido de fósforo del suelo, la calidad del ambiente y del agua de los curs...

  6. Relaciones agua-solutos-matriz insoluble en algunos sistemas alimentarios.

    OpenAIRE

    Chenoll Cuadros, Maria De La Creu

    2008-01-01

    Una gran parte de los alimentos estructurados pueden considerarse, de forma simplificada, constituidos por tres componentes: una matriz insoluble (que puede ser, por ejemplo, proteica, o compuesta por carbohidratos), agua y una serie de sólidos solubles (añadidos o no). Las relaciones de estos componentes entre sí, sus interacciones, y el modo en que varían durante determinados procesos industriales son factores que condicionan sus propiedades, tales como nutricionales y de calidad sensorial ...

  7. Reduced moderation BWR with advanced recycle system (BARS)

    International Nuclear Information System (INIS)

    Full text: The innovative concept for a reduced-moderation spectrum boiling water reactor (BWR) combined with an advanced fuel cycle system, which is named BARS, has been being developed as one of the IVNET (Innovative and Viable Nuclear Energy Technology) development projects from 2000 in Japan. The reduced-moderation spectrum BWR with high conversion property features higher utilization of uranium resources, multi-recycling usage of plutonium and flexibility in loading the extracted long-lived radioisotopes, it can realize a sustainable nuclear energy system without a liquid metal fast breeder reactor (LMFBR). The advanced recycle system means the combination of dry reprocessing and vibro-packing of MOX fuel fabrication. This system simplifies reprocessing and MOX fuel fabricating process and reduces related back end cost. The reduced moderation spectrum condition in BWR is obtained through triangular tight fuel rod lattice configuration and higher void fraction. This feature cause some technological problems; smaller void coefficients of positive side and larger uncertainty of nuclear and thermal hydraulic design method. According to the above problems, the developing project consists the three parts; core design study that enables the negative void coefficient, critical assembly experiment to validate nuclear design method and thermal hydraulic test to evaluate thermal hydraulic property. The core design study focuses on optimization of tight lattice bundle design. In the optimization of bundle design, it is important to satisfy both of the requirements for high conversion ratio and negative void coefficient. Streaming channel concept has been introduced to realize the negative void coefficient. The streaming channel consists of short bundle and upper vacant tube, which promotes neutron-streaming effect in higher void fraction. The neutron-streaming effect evaluated by Monte Carlo calculation showed the effectiveness of streaming channel concept for realizing

  8. ATRIUMTM Fuel - Continuous Upgrading for High Duty BWR Plants

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to Boiling Water Reactors worldwide, representing today more than 60 000 fuel assemblies. Since first delivered in 1992, ATRIUMTM10 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. Among them, the latest versions are ATRIUMTM 10XP and ATRIUMTM 10XM fuel assemblies which have been delivered to several utilities worldwide. During six years of operation experience reaching a maximum fuel assembly burnup of 66 MWd/kgU, no fuel failure of ATRIUMTM 10XP/XM occurred. Regular upgrading of the fuel assemblies' reliability and performance has been made possible thanks to AREVA NP's continuous improvement process and the 'Zero tolerance for failure' program. In this frame, the in-core behavior follow-up, manufacturing experience feedback and customer expectations are the bases for setting improvement management objectives. As an example, most fuel rod failures observed in the past years resulted from debris fretting and Pellet Cladding Interaction (PCI) generally caused by Missing Pellet Surface. To address these issues, the development of the Improved FUELGUARDTM debris filter was initiated and completed while implementation of chamfered pellets and Cr doped fuel will address PCI aspects. In the case of fuel channel bow issue, efforts to ensure dimensional stability at high burnup levels and under challenging corrosion environments have been done resulting in material recommendations and process developments. All the described solutions will strongly support the INPO goal of 'Zero fuel failures by 2010'. In a longer perspective, the significant trend in nuclear fuel operation is to increase further the discharge burnup and/or to increase the reactor power output. In the majority of nuclear power plants worldwide, strong efforts in power up-rating were made and are still ongoing. Most

  9. Tratamiento de aguas residuales textiles mediante un biorreactor de membrana

    Directory of Open Access Journals (Sweden)

    Lorena Salazar Gámez

    2009-01-01

    Full Text Available En la actualidad, los investigadores dedicados al tratamiento de agua residual se han esforzado por obtener mejores rendimientos de depuración, reducción de costos de operación y mantenimiento en la generación de procesos que posean una elevada flexibilidad para soportar las variaciones en el afluente, que generen una mínima producción de lodos y que sean diseñados en condiciones mínimas de área. En este trabajo se expone una nueva tecnología en el tratamiento de agua residual, la de los Biorreactores de Membrana (BRM, que nace como respuesta de estas exigencias, ya que combina el sistema de fangos activados (FA convencional y la filtración por membrana. Así mismo evalúa comparativamente el proceso de fangos activados y Biorreactor de Membrana a escala piloto en aguas de difícil tratamiento como son los efluentes textiles. Los resultados indican que el proceso de BRM reduce un 82 - 92% de la materia orgánica (DQO, un 95% de los sólidos en suspensión totales (SST, mientras que el proceso de fangos activados reduce un 54- 70% de la materia orgánica (DQO y un 32- 43% de SST.

  10. MICROCONTAMINANTES EMERGENTES EN AGUAS: TIPOS Y SISTEMAS DE TRATAMIENTO

    Directory of Open Access Journals (Sweden)

    Yolanda Patiño

    2014-01-01

    Full Text Available En el presente artículo se revisan las distintas clases de contaminantes emergentes, su origen y uso, y su presencia en las cuencas españolas, tanto en aguas superficiales como subterráneas. Los microcontaminantes emergentes son compuestos de distinto origen y naturaleza que hasta ahora habían pasado inadvertidos debido a su baja concentración (en el orden de las ppb y sobre los que no hay regulación específica. Están divididos en seis grandes grupos y muchos de ellos se comportan como disruptores endocrinos, provocando efectos negativos sobre la salud humana y el medio ambiente. Se encuentran presentes en las aguas ya que las plantas de tratamiento de aguas residuales no están diseñadas para su eliminación, por lo que están siendo vertidos. Se discuten las distintas alternativas para su eliminación – tratamiento físico-químicos, biológicos y tecnologías híbridas - , destacando dentro de los procesos físico-químicos el empleo de procesos avanzados de oxidación (AOPs que están resultando muy prometedores.

  11. Accumulation of operator workload data by using A-BWR training simulator

    International Nuclear Information System (INIS)

    Human-machine interface (HMI) of A-BWR has been developed in order to improve operational safety, reliability and to reduce workload. A-BWR HMI is fully computerized. JNES (Japan Nuclear Energy Safety Organization) and BTC (BWR Operator Training Center Corporation) have accumulated the operator workload quantitative data, related to the observation and operation at typical transient conditions, in order to evaluate the difference of operational workloads between A-BWR HMI and conventional type HMI. The workload evaluation shows the following results: - The workload density (observation and operation frequencies per unit time) of ABWR just after the plant trip is less than that of BWR-5. - At stable conditions after the transient, the workload density of ABWR becomes higher comparing that of BWR-5. A-BWR alarm system may increase the workload density caused by alarm multi-layer structure, because an operator has to use the flat and/or the CRT display to pursue every alarm. The analysis of shift team training at BTC shows that total workload is reduced at ABWR but alarm confirmation work still remains as burden. These results show some modifications might be needed for future HMI. To grasp the tendency of operator workload difference by the control panel type difference, the operator workload quantitative data have accumulated using ABWR type simulator and conventional type simulator at the same typical transient condition. These data were arranged operation frequency data, number of alarm generating data and number of switching CRT pictures data according to plant behaviour. The ABWR type HMI's characteristic have become clear by these operator workload data and the team characteristic evaluation data which BTC evaluated comparing the team performance difference of HMI type

  12. Calderón y las aguas revueltas de Guárdate del agua mansa

    Directory of Open Access Journals (Sweden)

    García Santo-Tomás, Enrique

    2004-04-01

    Full Text Available Guárdate del agua mansa es una de las piezas calderonianas de mayor actualidad. La comedia desarrolla el tema de las apariencias tratado desde una óptica bastante original, en donde el «agua mansa» alude a todas aquellas jóvenes que parecen no tener capacidad de acción, pero que al final son con quienes hay que estar más precavido. Sin embargo, la pieza también nos habla de un fascinante Madrid que seduce a la juventud ya desde el mismo inicio de la pieza, un Madrid visual que entra por los ojos y que conquista al instante.…

  13. The Japanese utilities' requirements for a next century BWR

    International Nuclear Information System (INIS)

    This paper reports on the progress of studies to establish a plant concept for a Boiling Water Reactor (BWR) of the next century. The studies were initiated in 1990 by the Japanese utilities, jointly with NSSS vendors, to investigate evolutionary and long term nuclear power plants. The plant concept is based on the evolution of the ABWR taking advantage of new technology. Fundamental plant philosophies are expressed by the following four desired characteristics: Economical, Benign to human, Simple, Flexible. According to these philosophies, concrete objectives of the plant design are reduction of operating burden and maintenance, increase of safety margin and flexibility to adjust to possible changes in economic circumstances in the years to come. The basic utilities' requirements for the new generation BWR were discussed based on the future social needs and the current operational experiences. Start of operation is to be in the 2010's when the early generation LWRs may need to be replaced. Plant power generation capacity will be about 1500 MWe since this level rating will be achievable by extrapolation of current technology. One important requirement is to achieve power generation costs competitive with other generation methods. An outline of the utilities' requirements follows: Operability; prevent inadvertent reactor scram and engineering safety system actuation due to single failure of normal duty systems or single operator error, achieve same load following capability as ABWR, design for plant availability of up to 90%, achieve plant design life of 60 years, maintain annual inspection period at less than 40 days, reduce maintenance activities in harsh environments, reduce employees' dose to less than that of ABWR, consider 'N+2' design to reduce peak loads during annual inspection. Safety margin; increase grace period for transient and accident events, adopt severe accident countermeasures, keep core damage frequency lower than that of ABWR and conditional

  14. Microorganismos indicadores de la calidad del agua potable en cuba

    Directory of Open Access Journals (Sweden)

    Marlen Robert Pullés

    2014-01-01

    Full Text Available Los agentes patógenos transmitidos por el agua constituyen un problema mundial que demanda un urgente control. Las bacterias, virus y parásitos causan enfermedades que varían en severidad. La determinación de microorganismos en el agua de consumo y su concentración proporcionan herramientas de control, indispensables para la toma de decisiones. Los controles rutinarios de todos los microorganismos, potencialmente riesgosos para la salud, resultan difíciles de llevar a cabo, debido a que ello representa, varios días de análisis y costos elevados. Para hacer una evaluación sencilla, económica y fiable de la presencia de patógenos, la vigilancia de la calidad del agua se efectúa mediante indicadores de contaminación, aplicando diferentes enfoques técnicos, cada uno con sus cualidades, defectos y limitaciones. El objetivo de esta revisión fue analizar el enfoque actual existente a nivel nacional e internacional, en relación con los indicadores microbiológicos empleados para la evaluación del agua potable, como elementos clave, y a partir de estos, proponer un esquema de monitoreo en Cuba. Los resultados reflejaron, la alternativa de considerar, la aplicación de un esquema para monitoreo complementario en Cuba, que incluya como indicadores de contaminación del agua potable además de las bacterias, algunos agentes biológicos no considerados en la norma, como los virus y los parásitos. Asimismo indicaron la necesidad de establecer un valor de referencia y definir los microorganismos a emplear en los monitoreos de validación, operativo o verificación. Esta propuesta aportaría importante información para la actualización de la norma cubana sobre la base del conocimiento de los estándares internacionales más reconocidos.

  15. Hydraulic modeling and simulation of a System Division of Essential Service Water in a BWR plant with Flow master; Modelo hidraulico y simulacion de una division del Sistema de Agua de Servicio Esencial de una central BWR con Flowmaster

    Energy Technology Data Exchange (ETDEWEB)

    Vegazo Juzgado, L.; Rodriguez Garcia, G. M.; Mota Coloma, M.

    2012-07-01

    At the conclusion of the project can say that Flow master is a simulation tool that allows you to create your model from a library of components and obtain useful results from the point of view of the operation, engineering and maintenance. Compared to previous software from the point of view of use, can comment that Flow master is a tool which has an intuitive and user-friendly interaction between the user and the program thus facilitating the modeling of the system and definition of the components of same.

  16. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B4C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  17. An A BWR demonstration simulator for training and developing technical staff

    International Nuclear Information System (INIS)

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

  18. Space–time convergence analysis on BWR stability using TRACE/PARCS

    International Nuclear Information System (INIS)

    Highlights: ► Quantify TRACE/PARCS space–time discretization error for BWR stability prediction. ► Establish space and time discretization necessary for space–time converged models. ► Show that the space–time converged model gives more reliable results for both stable and unstable reactor. ► Use of the space–time converged model increases confidence in the prediction of BWR stability. -- Abstract: Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.

  19. An A BWR demonstration simulator for training and developing technical staff

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  20. Upgrading BWR training simulators for annual outage operation training

    International Nuclear Information System (INIS)

    Based upon the recently developed quality assurance program by the Japanese electric companies, BWR Operator Training Center (BTC) identified the needs to enhance operators' knowledge and skills for operations tasks during annual outage, and started to develop a dedicated operator training course specialized for them. In this paper, we present the total framework of the training course for annual outage operations and the associated typical three functions of our full-scope simulators specially developed and upgraded to conduct the training; namely, (1) Simulation model upgrade for the flow and temperature behavior concerning residual heat removal (RHR) system with shutdown cooling mode, (2) Addition of malfunctions for DC power supply equipment, (3) Simulation model upgrade for water filling operation for reactor pressurization (future development). We have implemented a trial of the training course by using the upgraded 800MW full-scope training simulator with functions (1) and (2) above. As the result of this trial, we are confident that the developed training course is effective for enhancing operators' knowledge and skills for operations tasks during annual outage. (author)

  1. Recent experience and development of BWR fuel at NFI

    International Nuclear Information System (INIS)

    This paper describes the results of recent investigations by Nuclear Fuel Industries, Ltd. (NFI) conducted in cooperation with BWR electric power companies in Japan regarding high burnup fuel behavior, i.e. fuel cladding corrosion and hydrogen pickup, degradation of pellet thermal conductivity with burnup, and fission gas release. The authors confirmed by pool inspection that 9x9 assemblies irradiated up to 53 GWd/t, which is the maximum burnup in our experience, showed good performance without any harmful phenomena. With respect to the advanced Zr alloy HiFi, it was confirmed that HiFi retained high corrosion resistance and showed low hydrogen pick up and good mechanical properties after six cycles of irradiation. Regarding the high burnup fuel behavior, it was confirmed that the thermal behavior of the fuel, such as pellet thermal conductivity degradation and fission gas release behavior beyond 80 GWd/t, was stable in the extrapolation range of the burnup fuel behavior between about 60-70 GWd/t. In addition, a fuel performance analysis code developed by NFI was verified to predict the data measured beyond 80 GWd/t well. (author)

  2. Water injection system for turbine driven BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a water injection system of a turbine driven nuclear reactor for maintaining the function thereof even upon occurrence of a severe accident in a BWR type nuclear reactor. That is, the system comprises a differential pressure detection means for measuring a pressure difference between the downstream of a the turbine and a reactor container and an interrupting means for stopping the supply of steams to the turbine when the differential pressure exceeds a predetermined value. With such a constitution, when the pressure in the turbine driven water injection system is locally increased, the differential pressure detection means detects the differential pressure, to interrupt the supply of the steams to the turbine. Further, upon occurrence of a severe accident that a pressure in the reactor container is abnormally elevated, differential pressure is not caused between the downstream of the turbine and the reactor container. Accordingly, a protection function is not operated by the differential pressure detection means. Accordingly, injection of coolants to the reactor can be continued even upon loss of AC power source. (I.S.)

  3. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

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

  4. Development of BWR operator training simulator and training support systems

    International Nuclear Information System (INIS)

    This paper describes a BWR operator training simulator and training support systems that have been developed with the aim of providing support throughout operator training. The operator training simulator is needed in order to improve simulation fidelity and enlarge simulation scope. A 3-dimensional reactor core model has been developed in order to improve the understanding of operators respecting neutronics through realistic training. A severe accident model has been developed for training operators and technical support center teams respecting plant operation and for studying various phenomena. The severe accident is simulated by connecting the physical parameters continuously from the conventional model to the severe accident model. An emergency procedure guideline support system is adopted in order to improve efficiency of operation training for emergencies, since the emergency operation procedures are complicated and based on multiple parameter conditions. The operator training support system is also introduced so as to help training instructors to evaluate the operation and to give instructions to operators to improve operational accuracy. An instructor's burden is eased by automatically evaluating the operation errors based on signals of a simulator. The effects of these systems are evaluated and found to be effective in an actual training center and in engineers' examinations. (author)

  5. Development and implementation of a BWR digital feedwater control system

    International Nuclear Information System (INIS)

    EPRI and Northern States Power Company (NSP) realized that fault-tolerant digital technology could improve Feedwater Control System reliability and operations. The Digital Feedwater Control System (DFCS) is the first major fault-tolerant digital control application in a nuclear power plant in the US. The microprocessor-based controller replaced the analog controller in the feedwater control loop to improve performance and reliability of control including ease of maintenance and spare parts supply. The DFCS at Monticello plant has been in operation since July 1986 without any failure of the control system. This system replaced and upgraded the main and start-up analog controllers at Monticello BWR. It features automatic control, on-line signal validation, controller self-diagnosis, and fault tolerance. The dual-redundant hardware configuration minimizes spare parts availability problems. At a control room panel, operators select each feedwater valve's operating mode (one or three element control, manual, and so on) or set bias inputs for individual feed-water-valve demand. These and other features permit more exact feedwater control system tuning, improving feedwater control in all modes of plant operation. Signal validation using parity-space techniques isolated failed sensors and permit system switching to accurate sensors, thus avoiding outages. To ensure successful operation of the system, extensive verification and validation effect were conducted. These included design reviews, factory acceptance testing using simulation code, site acceptance testing using full-scale plant simulator, and pre-operational and operational testing at Monticello plant

  6. BWR [boiling water reactor] shutdown margin model in SIMULATE-3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) technical specifications require that the reactor be kept subcritical (by some prescribed margin) when at room temperature rodded conditions with any one control rod fully withdrawn. The design of an acceptable core loading pattern may require hundreds or thousands of neutronic calculations in order to predict the shutdown margin for each control rod. Direct, full-core, three-dimensional calculations with the SIMULATE-3 two-group advanced nodal code require 3 to 6 CPU min (on a SUN-4 workstation) for each statepoint/control rod that is computed. Such computing and manpower requirements may be burdensome, particularly during the early core design process. These requirements have been significantly reduced by the development of a fast, accurate shutdown margin model in SIMULATE-3. The SIMULATE-3 shutdown margin model achieves a high degree of accuracy and speed without using axial collapsing approximations inherent in many models. The mean difference between SIMULATE-3 one-group and two-group calculations is approximately - 12 pcm with a standard deviation of 35 pcm. The SIMULATE-3 shutdown margin model requires a factor of ∼15 less CPU time than is required for stacked independent two-group SIMULATE-3 calculations

  7. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  8. A New Method for Early Anomaly Detection of BWR Instabilities

    International Nuclear Information System (INIS)

    The objective of the performed research is to develop an early anomaly detection methodology so as to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The research utilizes a model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, which is used as a generator of time series data for anomaly detection at an early stage. The model captures critical nonlinear features of coupled thermal-hydraulic and nuclear reactor dynamics and (slow time-scale) evolution of the anomalies as non-stationary parameters. The time series data derived from this nonlinear non-stationary model serves as the source of information for generating the symbolic dynamics for characterization of model parameter changes that quantitatively represent small anomalies. The major focus of the presented research activity was on developing and qualifying algorithms of pattern recognition for power instability based on anomaly detection from time series data, which later can be used to formulate real-time decision and control algorithms for suppression of power oscillations for a variety of anticipated operating conditions. The research being performed in the framework of this project is essential to make significant improvement in the capability of thermal instability analyses for enhancing safety, availability, and operational flexibility of currently operating and next generation BWRs.

  9. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

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

  10. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

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

  11. BWR Fuel Lattice Design Using an Ant Colony Model

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose L.; Ortiz, Juan J. [Instituto Nacional de Investigaciones Nucleares, Depto. de Sistemas Nucleares, Carretera Mexico Toluca S/N. La Marquesa Ocoyoacac. 52750, Estado de Mexico (Mexico); Francois, Juan L.; Martin-del-Campo, Cecilia [Depto. de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico Paseo Cuauhnahuac 8532. Jiutepec, Mor. 62550 (Mexico)

    2008-07-01

    This paper deals with one of the steps of the nuclear fuel design: the radial fuel lattice design. It can be seen as a combinatorial optimization problem for determining the optimal 2D fuel rods enrichment and gadolinia distribution. In order to solve this optimization problem, the ant colony system technique is proposed. The main idea of the ant colony approach consists of emulating the real ant colony behaviour in their searching for minimum paths between two given points, usually between the nest and a food source. In this case, the environment where the artificial ants move is the space defined by the discrete possible values of Gd{sub 2}O{sub 3} contents, the U{sup 235} enrichment, and the valid locations inside the 10x10 BWR fuel lattice array. In order to assess any candidate fuel lattice in the optimization process, the HELIOS neutronic transport code is used. The results obtained in the application of the implemented model show that the proposed technique is a powerful tool to tackle this step of the fuel design. (authors)

  12. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the occasions: (1) confirming core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) measurements of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurement of isotopic compositions of fission product nuclides on high-burn up BWR UO2 fuels and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  13. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) analysis of the measurement data of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurements of isotopic compositions of fission product nuclides on high-burnup BWR UO2 fuels and the analysis of the measurement data, and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  14. Natural heat transfer augmentation in passive advanced BWR plants

    International Nuclear Information System (INIS)

    In the European Simplified Boiling Water Reactor (ESBWR), the long-term post-accident containment pressure is determined by the combination of non condensable gas pressure and steam pressure in the wet well gas space. Since there are no active systems for heat removal in the wet well, energy transmitted to the wet well gas space, by a variety of means, must be removed by passive heat transfer to the walls and suppression pool (SP). The cold suppression pool located below the hotter gas space provides a stable configuration in which convection currents are suppressed thus limiting heat and mass transfer between the gas space and pool. However, heat transfer to the walls results in natural circulation currents that can augment the heat and mass transfer to the pool surface. Using a simplified model, parametric studies are carried out to show that augmentation of the order of magnitude expected can significantly impact the heat and mass transfer to the pool. Additionally a review of available literature in the area of augmentation and mixed convection of this type is presented and indicates the need for additional experimental work in order to develop adequate models for heat and mass transfer augmentation in the configuration of a BWR suppression pool. (author)

  15. Quantitative evaluations of BWR regional oscillations using higher harmonics subcriticalities

    International Nuclear Information System (INIS)

    A quantitative study of a mechanism for BWR regional stability has been carried out from the higher harmonics viewpoint using a three-dimensional higher harmonics analysis code and is reported in this paper. The results show that the first azimuthal harmonics subcriticality takes a relatively small value under the regionally unstable condition. The calculated patterns of the first azimuthal harmonics agree quite well with the measured regional oscillation mode patterns. Comparing the subcriticality and the steady state power distribution, it is shown that the distribution exists whose first azimuthal harmonics subcriticality takes a small value. This implies that regionally unstable conditions can be estimated from fundamental mode distributions. The decomposition method of the oscillated power responses into the harmonics modes is presented. The results show that core-wide oscillation power responses almost entirely consist of the fundamental mode and regional oscillation power responses almost entirely consist of the first azimuthal harmonics mode. This indicates that the regional oscillation is the phenomenon in which the first azimuthal harmonics mode oscillates on the basis of the fundamental mode. (author)

  16. Impurity hideout/hideout return at the Susquehanna 2 BWR

    International Nuclear Information System (INIS)

    An impurity hideout return study was performed at the Susquehanna 2 BWR to provide an understanding of impurity hideout processes during normal operation and their impact on high temperature solution chemistry in corrosion product deposits on the fuel. Limited hideout return data obtained during shutdowns at 10 BWRs previously had indicated reasonable consistency with expectations based on MULTEQ high temperature solution chemistry modeling of hideout processes. Observations at Susquehanna 2 were consistent with expectations. Cumulative returns of species forming precipitates at low concentration factors above the bulk water concentration, e.g., calcium, magnesium, sulfate and silica were much greater than those of species having a minimal tendency to precipitate, e.g., sodium and chloride. Solutions present in the fuel cladding surface during normal operation were predicted to contain high concentrations (0.1 to 2 molal) of sodium, potassium, chloride, sulfate, silica and nitrate. The predicted solution pH at 300 degrees C was 9.4 (neutral pH = 5.5). The increase in conductivity observed during and after shutdown was shown to be due to solubilization of precipitates with retrograde solubilities rather than chemical/resin intrusion. Variations in reactor water concentrations during reactor water cleanup system isolation and power reductions were consistent with predictions developed from a mass balance around the reactor coolant system

  17. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

  18. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  19. Derivation of general scaling criteria for BWR containment tests

    International Nuclear Information System (INIS)

    General top-down scaling criteria for facilities used to study Boiling Water Reactor (BWR) containments including a pressure suppression system are derived, with particular attention to the recent passive BWRS. The criteria are derived by considering the generic processes in classes of containment subsystems (e.g., containment volumes, pools, pipes, etc.). In reactor containments, the thermodynamic behavior of the system (essentially, its pressure history) is linked to its thermal-hydraulic behavior (the flows of mass and energy between volumes). The case of prototypical fluids under prototypical thermodynamic conditions is treated. The study confirms the validity of the (familiar) scaling of power, volumes, horizontal areas in volumes, mass flow rates, and heat transfer areas with a system scale. Important pressure drops and the corresponding flows are controlled by the submergence depth of vents or by hydrostatic pressure differences in connected vessels. The analysis of these processes justify the choice of 1:1 scaling for the pressure drops, vertical heights, submergence depths and level differences. The importance of certain distortions regarding inertial response and transit times is minor

  20. Dry well cooling systems in BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Purpose: To prevent the damages of pipeways due to salt damages at the surface of control rod drives in BWR type reactors. Constitution: In control rod drives and the lowermost area in the dry well in which surface corrosion and pitching have been resulted by the salt contents in air due to the increase in the humidity accompanying the lowering of the temperature, a blower is disposed to the upstream of the cooling coils and a portion of high temperature air returned to the lower cooler is replaced with a low temperature feed air to increase the feed temperature in the area. Further, by upwardly turning the downwarded feed air drawing port in which cold feed air has so far been descended as it is, the descendance of the cold air is suppressed. As a result, temperature lowering in the driving mechanisms and the lower area can be prevented to obtain a predetermined temperature, whereby the dewing on the surface can be prevented and thereby preventing the occurrence of corrosion and pitching. (Horiuchi, T.)

  1. A damage tolerance analysis of an old BWR pressure vessel

    International Nuclear Information System (INIS)

    A program has been completed in Sweden where the oldest of the Swedish BWR-units has been subjected to an extensive investigation regarding signs of aging and degradation. The core and all the internal parts were removed and the stripped pressure vessel was decontaminated. Almost every weld in the vessel including all nozzles in the bottom head have then been inspected by aid of UT-inspection. An important part of the programs was the damage tolerance analysis. It has involved postulated surface cracks and embedded cracks (positioned along and across the welds) in all the inspected welds. By using the R6-method the maximum acceptable and critical crack size have then been determined for the most limiting load case and accounting for the individual material properties of each weld, cladding and base material. Of special interest is the core region. The base material is made of A 302 Grade B with rather high Cu- and Ni-contents, which have caused irradiation embrittlement in the beltline region. This implies that during certain cold loading cases, the critical crack size for a postulated surface crack in the core region, can be quite small. However, for load cases during normal operation, the material is on the upper shelf region and the critical crack sizes are large

  2. Design of a redundant meteorological station for a BWR reactor

    International Nuclear Information System (INIS)

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  3. Nonlinear behavior under regional neutron flux oscillations in BWR cores

    International Nuclear Information System (INIS)

    A three-dimensional time-domain core analysis code was applied to numerical simulations for an actual regional neutron flux oscillation observed in a commercial BWR core, in order to investigate potential nonlinear behavior in its coupled neutronic and thermohydraulic system. The present study shows existence of the nonlinear reactivity interaction between the fundamental and first azimuthal spatial harmonics modes of neutron flux distribution under the regional event. The spectrum analysis of the simulated data provides a unique result, that is, temporal harmonics peaks are excited at the even- and odd-order multiples of the characteristic resonance frequency in the fundamental and first spatial harmonics responses, respectively. The numerical simulation also shows that the strong nonlinearity of the coupled neutronic and thermohydraulic dynamics locally appears where the power unstably oscillates with large amplitudes, inducing the power shift and reactivity bias which are shown in the core-wide situation under the global oscillations. This contributes to suppression of the divergence of the local power oscillation, and also to development of the saturated self-limited cycles under the regional oscillations. (author)

  4. BWR Fuel Lattice Design Using an Ant Colony Model

    International Nuclear Information System (INIS)

    This paper deals with one of the steps of the nuclear fuel design: the radial fuel lattice design. It can be seen as a combinatorial optimization problem for determining the optimal 2D fuel rods enrichment and gadolinia distribution. In order to solve this optimization problem, the ant colony system technique is proposed. The main idea of the ant colony approach consists of emulating the real ant colony behaviour in their searching for minimum paths between two given points, usually between the nest and a food source. In this case, the environment where the artificial ants move is the space defined by the discrete possible values of Gd2O3 contents, the U235 enrichment, and the valid locations inside the 10x10 BWR fuel lattice array. In order to assess any candidate fuel lattice in the optimization process, the HELIOS neutronic transport code is used. The results obtained in the application of the implemented model show that the proposed technique is a powerful tool to tackle this step of the fuel design. (authors)

  5. Identification of the reduced order models of a BWR reactor

    International Nuclear Information System (INIS)

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  6. A New Methodology for Early Anomaly Detection of BWR Instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, K. N.

    2005-11-27

    The objective of the performed research is to develop an early anomaly detection methodology so as to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The research utilizes a model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, which is used as a generator of time series data for anomaly detection at an early stage. The model captures critical nonlinear features of coupled thermal-hydraulic and nuclear reactor dynamics and (slow time-scale) evolution of the anomalies as non-stationary parameters. The time series data derived from this nonlinear non-stationary model serves as the source of information for generating the symbolic dynamics for characterization of model parameter changes that quantitatively represent small anomalies. The major focus of the presented research activity was on developing and qualifying algorithms of pattern recognition for power instability based on anomaly detection from time series data, which later can be used to formulate real-time decision and control algorithms for suppression of power oscillations for a variety of anticipated operating conditions. The research being performed in the framework of this project is essential to make significant improvement in the capability of thermal instability analyses for enhancing safety, availability, and operational flexibility of currently operating and next generation BWRs.

  7. 75 FR 21034 - Notice of Availability of Record of Decision for the Agua Fria National Monument and Bradshaw...

    Science.gov (United States)

    2010-04-22

    ... Bureau of Land Management Notice of Availability of Record of Decision for the Agua Fria National... Agua Fria National Monument and Bradshaw-Harquahala Planning Area, located in central Arizona. The... occupied or used portions of the planning area during prehistoric or historic times. The Agua Fria...

  8. Calidad de las aguas superficiales en la Cuenca del Ebro. Tendencias de salinidad, nitrato y fosfato

    OpenAIRE

    Quílez Sáez de Viteri, Dolores

    2011-01-01

    Objetivo General: Caracterización de la calidad de las aguas superficiales de la Cuenca del Ebro y de sus tendencias. Redes de Control de Calidad de Aguas de CHE: - 497 estaciones de control - Datos instantáneos con frecuencias variables - Se han seleccionado 31 estaciones - Relevancia dentro de la Cuenca del Ebro - Predominio del regadío - Disponibilidad de datos

  9. LOCALIZACIÓN Y AFECTACIÓN DE COSTES MEDIOAMBIENTALES Y NO MEDIOAMBIENTALES EN LAS EMPRESAS DE AGUAS: EL COSTE DEL AGUA DESALADA

    Directory of Open Access Journals (Sweden)

    Jerónimo Pérez Alemán

    2005-12-01

    Full Text Available La Directiva 2000/60/EC del Parlamento Europeo y del Consejo establece un marco de acción comunitaria en el ámbito de la política de aguas. Esta norma indica, en su artículo 9, que los miembros de los Estados tendrán en cuenta el principio de recuperación de los costes de los servicios de aguas, incluidos los medioambientales. Así pues, las empresas de tratamiento y abastecimiento urbano de aguas tienen la necesidad de generar información sobre los costes en general, y sobre los costes (y otras magnitudes medioambientales en particular. En el presente trabajo centramos nuestra atención en el coste de uno de los outputs de tales empresas: el agua desalada.

  10. Design of a redundant meteorological station for a BWR reactor; Diseno de una estacion meteorologica redundante para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, R.; Celis del Angel, L.; Bucio, F.; Rivero, T.; Palacios, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: ramses@nuclear.inin.mx

    2008-07-01

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  11. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: augusto@correo.unam.mx

    2004-07-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

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

    International Nuclear Information System (INIS)

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

  13. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  15. Design guideline to prevent the pipe rupture by combustion of radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005, and the 2nd edition in March 2007. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2010, JANTI published the 3rd edition of the guideline. This is the report of the final edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent pipe rupture accident due to combustion of radiolysis gas. (author)

  16. Report on the BWR owners group radiation protection/ALARA Committee

    International Nuclear Information System (INIS)

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

  17. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

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

    International Nuclear Information System (INIS)

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

  19. La fundamentalidad del derecho al agua en Colombia

    Directory of Open Access Journals (Sweden)

    Mies Sutorius

    2015-12-01

    Full Text Available El agua es un recurso natural finito de vital importancia para el ser humano, lo que explica que con el paso del tiempo se busque con mayor fuerza su protección, no solo en el ámbito internacional sino también en el orden jurídico interno. De ahí el interés en realizar una revisión normativa y especialmente jurisprudencial que permita evidenciar la evolución del derecho al agua, los matices y las diversas formas de protección del mismo, profundizando en su dimensión personal. La Constitución Política colombiana omite enunciar el derecho al agua como derecho individual, razón por la cual se procede a estudiar la jurisprudencia constitucional, seleccionándose algunas sentencias en donde existen sujetos de especial protección, en tanto al reconocerse a estos la vulneración y garantía de derechos se prevé la protección de los derechos del resto de la población; así como algunos casos en donde existen condiciones particulares, a partir de la suspensión del servicio por falta de pago a la empresa encargada del suministro, conexión fraudulenta y reconexión ilegal; para concluir que se han presentado dos tendencias en el amparo de este derecho, que ha pasado de ser protegido por vía de la “teoría de la conexidad” a ser catalogado como fundamental con una tutela restringida a la cantidad mínima necesaria para sobrevivir.

  20. Acuaporinas: proteínas mediadoras del transporte de agua.

    Directory of Open Access Journals (Sweden)

    Julio César Sánchez

    2009-11-01

    Full Text Available Introducción. Todos los organismos vivos están compuestos en su mayoría por agua, siendo ésta fundamental para la homeostasis celular a todo nivel. Por esta razón su transporte a través de las membranas biológicas ha sido siempre un campo de gran interés en la fisiología. La investigación acerca de este tópico se ha incrementado notablemente en los últimos años a partir del descubrimiento de las acuaporinas, las cuales han permitido comprender mejor los mecanismos que la célula utiliza para el control de los flujos de agua a través de la membrana y por ende, la regulación de su osmolaridad interna. Tema general. Hasta el momento se han descrito 11 subtipos de acuaporinas (AQP0-10, las cuales comparten similitudes estructurales y se han relacionado con una gran diversidad de enfermedades en diversos sistemas, desde cataratas hasta diabetes insípida. Objetivo. Se presenta una visión del estado actual del conocimiento acerca de estas importantes proteínas y su relación con diversos procesos fisiológicos y patológicos. Conclusiones. Las acuaporinas son proteínas mediadoras del transporte de agua y las alteraciones en su funcionamiento pueden conducir a una gran diversidad de enfermedades, por lo cual deben ser un objeto prioritario de investigación en el futuro próximo para ayudar a comprender mejor su fisiopatología.

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

    International Nuclear Information System (INIS)

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

  2. Estudio del mecanismo de falla de terraplenes debido a la infiltración de aguas lluvias mediante el monitoreo de presiones de poros y contenidos de agua

    Directory of Open Access Journals (Sweden)

    EDWIN GARCÍA

    2007-01-01

    Full Text Available El presente artículo muestra experimentos realizados utilizando modelos a escala para estudiar el proceso de infiltración en terraplenes sometidos a aguas lluvias. Varios modelos a escala fueron construidos usando un suelo arenoso con alto contenido de finos; para aplicar la lluvia artificial y observar el proceso de infiltración del agua se utilizó un tubo de irrigación. Para monitorear los cambios en succión ocasionados por los ciclos de humedecimiento y secado se instalaron dentro del suelo varios sensores de presión de poros, contenido de agua y de desplazamiento en diferentes puntos. El cambio de la presión de poros en el tiempo, el contenido de agua y las relaciones entre las mediciones de presión de poros y desplazamientos del suelo son presentados. Este estudio muestra que el movimiento del suelo en un talud está directamente relacionado con el contenido de agua del suelo y las presiones de poros existentes en él, sugiriendo que por medio del monitoreo de estos dos parámetros es posible predecir fallas locales inducidas por infiltración de aguas lluvias en los taludes.

  3. Uso de plantas acuáticas en el tratamiento de agua y aguas residuales en la Universidad de Matanzas "Camilo Cienfuegos"

    Directory of Open Access Journals (Sweden)

    Damarys Santana Diaz

    2011-05-01

    Full Text Available En el presente trabajo  se analiza el uso de las plantas acuáticas para purificar el agua que se utiliza con fines ornamentales en el Parque de La Juventud de la Universidad de Matanzas "Camilo Cienfuegos", así como para el tratamiento de aguas residuales de dicha instalación. El estudio se realiza a escala de laboratorio y en ambos casos la efectividad del tratamiento se valora a partir de la determinación de parámetros como: demanda química de oxígeno, conductividad, pH, dureza total, dureza de calcio, dureza de magnesio, sólidos totales, fijos y volátiles y coliformes totales y fecales. Se comprobó que las plantas acuáticas pueden disminuir los valores de parámetros indicadores de la calidad del agua a niveles que permitan disminuir la frecuencia con que se cambia el agua de los espejos. Los resultados evidencian que el uso de la planta acuática Nymphaea ampla constituye una alternativa para tratar las aguas y mejorar su calidad con un efecto económico considerable por concepto de ahorro de agua potable.

  4. Matemáticas para la industria del agua

    OpenAIRE

    Izquierdo Sebastián, Joaquín; Pérez, Rafael; Vicente S. Fuertes; Iglesias, Pedro L.; López, Amparo

    2004-01-01

    En el campo del agua existe una enorme diversidad de actividades e intereses y, por tanto, de áreas de trabajo. Los problemas que se plantean en estas áreas son auténticos problemas de ingeniería y, como consecuencia, las ayudas que ciertas técnicas de Matemática Aplicada pueden prestar son realmente importantes. Por un lado, es preciso disponer de herramientas de análisis que permitan realizar simulaciones fiables de los distintos modelos que se plantean analizando diversas configuraciones, ...

  5. La electrolisis salina para el tratamiento de aguas de lastre

    OpenAIRE

    Mesa Salvany, Daniel

    2014-01-01

    El objetivo principal del proyecto, es el desarrollo de equipos electrolíticos para el tratamiento de aguas de lastre, acorde con los requerimientos de la normativa, y los del sector. Para ello, los objetivos específicos, son:  Definición de los requerimientos de la tecnología, teniendo en consideración todas las normativas navales que afecten al proyecto.  Definición de los requerimientos de la tecnología, teniendo en consideración todas las especificaciones técnicas y comerciales qu...

  6. Hormigón en aguas y terrenos agresivos

    Directory of Open Access Journals (Sweden)

    García de Paredes y Gaibrois, Pablo

    1962-09-01

    Full Text Available Not availableEl Dr. Ingeniero H. J. Wierig, del Laboratorio de la Industria del Cemento de Westfalia, ha publicado, en la revista «Bau und Bauindustrie» (núm. 6 de 1961, un interesante artículo acerca de la Preparación del hormigón en aguas y suelos agresivos, de~ cual entresacamos las siguientes notas con la complacencia que supone encontrar totalmente coincidentes, las opiniones del Dr. Wierig con el contenido de las «Recomendaciones» publicadas en la Instrucción sobre hormigón armado H. A. 61 de este Instituto.

  7. Desplazamiento de aceite por agua en medios porosos fracturados vugulares

    Directory of Open Access Journals (Sweden)

    Juana Cruz

    2001-01-01

    Full Text Available Se presentan experimentos llevados a cabo con una celda bidimensional de acrílico, que presenta un medio poroso fracturado vugular. Esta celda tiene seis vúgulos conectados con fracturas, un pozo de inyección y uno de producción. Los primeros resultados indican que el modelo explica de manera confiable los mecanismos que están presentes en los procesos de inyección de agua en yacimientos naturalmente fracturados, en presencia de vúgulos.

  8. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

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

  9. Control Rod Pattern Planning of a BWR using Enhanced Nelder-Mead Method

    International Nuclear Information System (INIS)

    We propose a new optimization algorithm for the short-term planning of control rod patterns in an operating BWR. This algorithm is based on the enhanced Nelder-Mead simplex method in which convergence ability is improved for constrained problems in several ways. The main characteristic of this approach is it uses continuous values for the axial positions of control rods. Through calculations in an actual BWR plant, we showed that the new algorithm is effective for automation of short-term planning and reduction of the engineer's workload. (authors)

  10. Validation and application of TRACE for transient analysis of a German BWR plant

    International Nuclear Information System (INIS)

    The Karlsruhe Institute of Technology (KIT) is participating on CAMP to validate TRACE for LWR transient analysis. The validation of safety-relevant heat transfer models of TRACE, data BFBT bundle tests are available for institutions taking part on international OECD/NEA Benchmarks. In the first step, the BWR-relevant models of TRACE were validated using experimental data. Then, an integral plant model of a German BWR of Type-72 was developed using the multidimensional VESSEL-component of TRACE to simulate a plant events such as the TUSA. In this paper, the validation work, the plant modeling and selected results will be presented and discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  12. Interpretation of cross-correlation transit time from LPRM signals in a BWR

    International Nuclear Information System (INIS)

    In a commercial BWR, the cross correlation between two noise signals from Local Power Range Monitors is performed. As a consequence, a transit time is determined. There is not yet a wide acceptation of what exactly this transit time means. Four different techniques for the noise analysis are used; the results were alike, but the Rehocence method stood out as the most robust. Noise measurements were performed in a commercial BWR. Understanding what the transit time means, an attempt is made. Comparison with the output of a detailed thermohydraulic code shows agreement with the transit time measurements. (authors)

  13. The Delft desire facility for studies on (natural circulation) BWR primary system statics and dynamics

    International Nuclear Information System (INIS)

    A test facility for research on BWR core statics and dynamics was designed and built in Delft. The loop, DESIRE, consists of a BWR fuel assembly, a riser, condenser and a downcorner section. Freon-12 is used as a coolant. Presently, research on this facility is focused on investigations of the physical aspects of natural-circulation cooling and reactor kinetic stability. To this end, an artificial feedback from in-core void fraction to heating power is being established. The void fraction is determined on a sub-channel level by measuring the transmission of a collimated gamma beam

  14. Generic safety evaluation report regarding integrity of BWR scram system piping

    International Nuclear Information System (INIS)

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

  15. Recursos de agua no convencionales en España: estado de la cuestión, 2010

    OpenAIRE

    Olcina Cantos, Jorge; Moltó Mantero, Enrique

    2010-01-01

    Los recursos de agua no convencionales han experimentado un notable desarrollo durante la primera década del presente siglo. El cumplimiento de la Directiva 91/271 de depuración de aguas residuales urbanas y la puesta en marcha del programa AGUA han resultado decisivos para el impulso de la depuración y desalación de aguas en España. Por el contrario, la reutilización de aguas residuales depuradas, aunque cuenta con actuaciones significativas (agricultura, usos recreativos), sigue siendo una ...

  16. Planificación y gestión del agua subterránea: aspectos generales y retos

    OpenAIRE

    Custodio Gimena, Emilio

    2011-01-01

    El agua subterránea, además de ser esencial para el medio ambiente, es un importante recurso y reserva de agua para satisfacer las necesidades humanas. Por ello es susceptible de extracción intensiva, cosa que ya sucede en numerosos acuíferos a nivel mundial. La extracción de agua subterránea modifica el ciclo natural del agua, lo que puede originar importantes cambios en la Naturaleza y alterar las relaciones con las otras fases del ciclo hidrológico. La explotación del agua subterránea p...

  17. Timing criteria for supplemental BWR emergency response equipment

    International Nuclear Information System (INIS)

    The Great Tohuku Earthquake and subsequent Tsunami represented a double failure event which destroyed offsite power connections to Fukushima-Daiichi site and then destroyed on-site electrical systems needed to run decay heat removal systems. The accident could have been mitigated had there been supplemental portable battery chargers, supplemental pumps, and in-place piping connections to provide alternate decay heat removal. In response to this event in the USA, two national response centers, one in Memphis, Tennessee, and another in Phoenix, Arizona, will begin operation. They will be able to dispatch supplemental emergency response equipment to any nuclear plant in the U.S. within 24 hours. In order to define requirements for supplemental nuclear power plant emergency response equipment maintained onsite vs. in a regional support center it is necessary to confirm: (a) the earliest time such equipment might be needed depending on the specific scenario, (b) the nominal time to move the equipment from a storage location either on-site or within the region of a nuclear power plant, and (c) the time required to connect in the supplemental equipment to use it. This paper describes an evaluation process for a BWR-4 with a Mark I Containment starting with: (a) severe accident simulation to define best estimate times available for recovery based on the specific scenario, (b) identify the key supplemental response equipment needed at specific times to accomplish recovery of key safety functions, and (c) evaluate what types of equipment should be warehoused on-site vs. in regional response centers. (authors)

  18. Description and status of BWR water radiolysis model

    International Nuclear Information System (INIS)

    A water radiolysis model for boiling water reactor primary coolant circuits has been developed jointly with Harwell Laboratories. Hydrogen water chemistry plant tests, on which the model is based, started in 1982. Since then, ten plant tests have been completed. Test results indicate that: (1) oxygen concentrations in the steam from plant to plant decreases fairly uniformly with increasing hydrogen concentration; (2) oxygen concentrations in the recirculating water varies widely with increasing hydrogen concentration in the feedwater; and (3) copper significantly suppresses the reduction of oxygen. The radiolysis model was developed to provide predictive conditions in the BWR primary circuit including the core. The objectives of the work are: (1) to develop and calibrate the model against experimental data; (2) to predict the concentration of various radiolytic species and oxygen suppression in various parts of the primary system under hydrogen water chemistry conditions; and (3) to define the role of copper on the reactions. The model considers the thermodynamics and kinetics of homogeneous chemical reactions in eight regions of the reactor and its external circuit. From initial concentrations, the model constructs a set of simultaneous equations to calculate hydrogen, hydrogen peroxide, and oxygen concentrations in various regions. Pilgrim data were used for the initial calibration. The model includes neutron and gamma dose rates in the core region and gamma dose rates in the other various regions, as well as temperature, power levels, heat transfer, steam quality, and gas/liquid transport in the core region, etc. The results of calculations for the oxygen concentrations in the recirculation water as a function of hydrogen concentration in the feedwater show good agreement with the experimental data obtained in a number of reactor tests

  19. Liquid films and droplet deposition in a BWR fuel element

    International Nuclear Information System (INIS)

    In the upper part of boiling water reactors (BWR) the flow regime is dominated by a steam-water droplet flow with liquid films on the nuclear fuel rod, the so called (wispy) annular flow regime. The film thickness and liquid flow rate distribution around the fuel rod play an important role especially in regard to so called dryout, which is the main phenomenon limiting the thermal power of a fuel assembly. The deposition of droplets in the liquid film is important, because this process sustains the liquid film and delays dryout. Functional spacers with different vane shapes have been used in recent decades to enhance droplet deposition and thus create more favorable conditions for heat removal. In this thesis the behavior of liquid films and droplet deposition in the annular flow regime in BWR bundles is addressed by experiments in an adiabatic flow at nearly ambient pressure. The experimental setup consists of a vertical channel with the cross-section resembling a pair of neighboring subchannels of a fuel rod bundle. Within this double subchannel an annular flow is established with a gas-water mixture. The impact of functional spacers on the annular flow behavior is studied closely. Parameter variations comprise gas and liquid flow rates, gas density and spacer shape. The setup is instrumented with a newly developed liquid film sensor that measures the electrical conductance between electrodes flush to the wall with high temporal and spatial resolution. Advanced post-processing methods are used to investigate the dynamic behavior of liquid films and droplet deposition. The topic is also assessed numerically by means of single-phase Reynolds-Averaged-Navier-Stokes CFD simulations of the flow in the gas core. For this the commercial code STAR-CCM+ is used coupled with additional models for the liquid film distribution and droplet motion. The results of the experiments show that the liquid film is quite evenly distributed around the circumference of the fuel rods. The

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  1. El cultivo de alfalfa utilizando agua de perforación, agua residual urbana y precipitaciones The cultivation of alfalfa using artesian well water, urban wastewater and rainfall

    Directory of Open Access Journals (Sweden)

    José O. Plevich

    2012-12-01

    Full Text Available El objetivo de este trabajo fue analizar la producción de biomasa aérea, la eficiencia del uso de agua y el valor nutritivo de alfalfa cultivada aplicando riego con agua residual urbana, agua de perforación y un testigo (precipitaciones, en el Campus de la Universidad Nacional de Rio Cuarto (Córdoba-Argentina. Los tratamientos de riego tuvieron un efecto positivo sobre la producción de biomasa aérea obteniéndose un 24% más de producción de biomasa que en la situación de secano (precipitaciones. Entre los tratamientos con riego también se manifestaron diferencias estadísticamente significativas. La alfalfa regada con aguas residuales urbana supero en un 19% a la cultivada con agua de perforación. La eficiencia del uso del agua por parte del cultivo se incrementó cuando el riego se realizó con aguas residuales urbanas. Además, se determinó que la alfalfa regada con agua residual, supera los valores nutritivos del cultivo que crece en condiciones sin riego; encontrándose 39% más de proteínas, 14% más de digestibilidad e igual porcentaje de energía metabólica. Se puede concluir que el agua residual urbana representa otra posible fuente de agua, factible de ingresar al suelo y estar a disposición para el cultivo de alfalfa.The objective of this study was to analyse the production of alfalfa biomass, the efficiency of water use and the nutritional value of forage, inside an irrigation structure of an experimental facility that uses treated urban wastewater, artesian well water and rainfall, at the campus of the Universidad National de Rio Cuarto (Córdoba-Argentina. The irrigation treatment had a positive effect on the production of biomass of alfalfa. On an average 24% increase in production was observed when compared to dry-land farming. Regarding the quality of the irrigation water, statistical differences in production were observed, the irrigation with urban wastewater produced 19% more than that of well water. The

  2. EFECTOS REDISTRIBUTIVOS DE LAS TARIFAS DE AGUA POTABLE

    Directory of Open Access Journals (Sweden)

    VÉLEZ GOMEZ LUIS DIEGO

    2010-05-01

    Full Text Available Aunque para Colombia la situación no es tan dramática, se avizora un problema de cobertura y escasez de agua potable para los niveles más desprotegidos de la población, en razón de su incapacidad de pago. La consecuencia inmediata es una acentuación de los índices de pobreza y de su manifestación más clara: el hambre. Este trabajo pretende mostrar que los niveles de pobreza son muy sensibles al precio de los servicios públicos domiciliarios, entre ellos, del agua, y atribuye a las fallas de inclusión y exclusión de los subsidios parte del grueso número de personas que hoy están por debajo de la línea de pobreza, aunque ésta no sea la única causa. Dada, entonces, la prominencia de las empresas de servicios públicos en la economía, los gobiernos tienen la obligación de resolver parte del difícil problema de equidad a través de las tarifas, sin dejar de garantizar la sostenibilidad económica de los proveedores del recurso.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  4. El abastecimiento de agua en las ciudades del Mediterráneo

    Directory of Open Access Journals (Sweden)

    Pulido Bosch, Antonio

    1999-10-01

    Full Text Available Not available.En las ciudades costeras, como en las no costeras, una buena parte del agua captada se pierde en las redes de distribución; además, el usuario no es siempre consciente de la necesidad de ahorrar al máximo este recurso. Hay que tener en cuenta igualmente que el agua residual, una vez depurada, puede constituir un buen recurso que permite satisfacer parte de la demanda y a la vez reducir los problemas planteados por la eliminación de las aguas usadas directamente en el mar. Las soluciones a los problemas de la escasez de agua deben pasar por el ahorro de agua, la reducción de las pérdidas en las redes de distribución, la reutilización y, a veces, la importación de agua de lugares excedentarios, sin olvidar la desalación de las aguas saladas o salobres. El desarrollo sostenible exige una planificación partiendo de la demanda máxima admisible, compatible con el respeto a los equilibrios ambientales; esto puede exigir la limitación del crecimiento de las ciudades, lo que no es siempre aceptado. Por último, se comentan dos ejemplos, uno representativo de una ciudad donde no existe gran competencia por el agua entre la demanda agrícola y la demanda urbana (Almería; y el otro relativo a una ciudad donde el turismo de verano aumenta casi diez veces la población, con una agricultura para la que la demanda de agua alcanza en verano también su máximo (Almuñécar, Granada.

  5. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  6. Development and Testing of CTF to Support Modeling of BWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gosdin, Chris [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

    2016-01-29

    This milestone supports developing and assessing COBRA-TF (CTF) for the modeling of boiling water reactors (BWRs). This is achieved in three stages. First, a new preprocessor utility that is capable of handling BWR-specic design elements (e.g., channel boxes and large water rods) is developed. A previous milestone (L3:PHI.CTF.P12.01) led to the development of this preprocessor capability for single assembly models. This current milestone expands this utility so that it is applicable to multi-assembly BWR models that can be modeled in either serial or parallel. The second stage involves making necessary modications to CTF so that it can execute these new models. Specically, this means implementing an outer-iteration loop, specic to BWR models, that equalizes the pressure loss over all assemblies in the core (which are not connected due to the channel boxes) by adjusting inlet mass ow rate. A third stage involves assessing the standard convergence metrics that are used by CTF to determine when a simulation is steady-state. The nal stage has resulted in the implementation of new metrics in the code that give a better indication of how steady the solution is at convergence. This report summarizes these eorts and provides a demonstration of CTF's BWR-modeling capabilities. CASL-U-2016-1030-000

  7. Analysis of relationship between stability and flow parameters in a BWR

    International Nuclear Information System (INIS)

    Results of quantitative analysis of mutual relationship between the BWR stability and channel steam velocity are presented. The stability parameter, defined by the damping ratio, and the steam velocity are estimated by analysis of neutron noise data from local power range monitor (LPRM) detector signals. These parameters are treated as varying randomly as a function of time

  8. Coupled field effects in BWR stability simulations using SIMULATE-3K

    International Nuclear Information System (INIS)

    The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17

  9. Predictions by the proper orthogonal decomposition reduced order methodology regarding non-linear BWR stability

    International Nuclear Information System (INIS)

    Unexpected non-linear boiling water reactor (BWR) instability events in various plants, e.g. LaSalle II in 1988 and Oskarshamn II in 1990 amongst others, emphasize the major safety relevance and the existence of parameter regions with unstable behavior. A detailed description of the complete dynamical non-linear behavior is of paramount importance for BWR operation. An extension of state-of-the-art methodology towards a more general stability description, also applicable in the non-linear region, could lead to a deeper understanding of non-linear BWR stability phenomena. With the intention of a full non-linear stability analysis of the two-phase BWR system, the present paper aims at a general non-linear methodology capable to achieve reliable and numerical stable reduced order models (ROMs), representing the dynamical behavior of an original system based on a small number of transients. Model-specific options and aspects of the proposed methodology are focused on and illustrated by means of a strongly non-linear dynamical system showing complex oscillating behavior. Prediction capability of the proposed methodology is also addressed. (orig.)

  10. Validation of BWR advanced core and fuel nuclear designs with power reactor measurements

    International Nuclear Information System (INIS)

    Power reactor measurements have been important in validating the reliability, performance characteristics and economics of BWR advanced core and fuel designs. Such measurements go beyond the data obtainable from normal reactor operation and provide detailed benchmark data necessary to verify design and licensing computer design and simulation models. In some cases, such as in the validation of the performance of zirconium barrier pellet-cladding-interaction (PCI) resistant cladding, the BWR power reactor measurements have subjected the advanced fuel design to operating conditions more severe than normal operating conditions, thereby providing nuclear-thermal-mechanical-corrosion performance data for accelerated or extended conditions of operation. In some cases destructive measurements have been carried out on BWR power reactor fuel to provide microscopic and macroscopic data of importance in validating design and licensing analysis methods. There is not uniform agreement among core and fuel designers on the needs for special power reactor core and fuel measurements for validation of advanced designs. The General Electric approach has been to error on the side of extensive, detailed measurements so as to assure reliable performance licensing and economic design and predictive capability. This paper is a summary of some of the validative power reactor measurements that have been carried out on advanced BWR core and fuel designs. Some comparisons of predictions with the data are summarized

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

    International Nuclear Information System (INIS)

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

  12. Boron concentration evolution in the temporary curtains of a BWR reactor. Burcur code

    International Nuclear Information System (INIS)

    The theoretical model and the user's guide of the code Burcur is included. This code analyzes the burnable poison concentration of the temporary curtains as a function of time, for BWR reactors of the 7 x 7 design. The computing time being reasonably short, the number of burnup steps is as high as necessary.(author)

  13. Bivariate empirical mode decomposition applied to the estimation of out-of-phase oscillations in BWR

    International Nuclear Information System (INIS)

    Highlights: • Bivariate empirical mode decomposition (BEMD) in BWR’s instabilities is studied. • The phase determines out-of-phase oscillations in the BWR instability. • The method based in BEMD does not represent a high computational complexity. • The methodology was validated with Nuclear Power Plants stability benchmarks. • The results show that the method contributes to detect out-of-phase oscillations. - Abstract: In this paper a new method based on the bivariate empirical mode decomposition to estimate the phase of regional (out-of-phase) or global (in-phase) modes associated with instabilities in boiling water reactors (BWR), is explored. The proposed method allows decomposing the analyzed signal (constructed from two different Local Power Range Monitors, LPRMs) in different levels or intrinsic mode functions (IMF). The estimation of the phase between these LPRM signals can be achieved by tracking the modes associated to the instability of the BWR and obtaining the cross-correlation function of their corresponding IMF. This phase determines possible out-of-phase oscillations, which play an important role in the BWR instability. The method is relatively simple to implement and it does not represent a high computational complexity. The methodology was tested with simulated signals and validated with two events reported in the Forsmark and Ringhals stability benchmarks. The results of the cases studied show that the proposed method clearly contributes on the fact to detect possible cases of out-of-phase oscillations

  14. Technical report on material selection and processing guidelines for BWR coolant pressure boundary piping

    International Nuclear Information System (INIS)

    This report sets forth the NRC staff's revised acceptable methods to reduce the intergranular stress corrosion cracking susceptibility of BWR ASME Code Class 1 and 2 pressure boundary piping and safe end. For plants that cannot fully comply with the material selection, testing, and processing guidelines of this document, varying degrees of augmented inservice inspection and leak detection requirements are presented

  15. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  16. Water chemistry management of nuclear power plant. Water chemistry management of BWR plant

    International Nuclear Information System (INIS)

    There are two kinds of nuclear power plants such as Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) in Japan. In this paper, a water chemistry management of BWR plant is explained. BWR plant makes steam produced in the reactor send to the turbine and produce power, then condensate in the main condenser and use again as feed water. The objects of water chemistry management of BWR are security of good conditions of fuel and structure materials and reduction of the dose equivalent and the radioactive waste. The volume of coolant depends on the temperature change, the concentration of boric acid for neutron absorber, lithium hydroxide for pH control and hydrogen gas for corrosion are controlled. Impurity metals in water of reactor are removed by the condensate demineralizer. The concentration of boron and lithium is controlled from 0 to 4000 ppm and from 0.2 to 2.2 ppm, respectively. On water chemistry technologies for dose reduction, oxygen injection into feed water and control operation of rate of Ni/Fe are explained. On the technologies for preventive maintenance, degassing operation of reactor and hydrogen injection into feed water are described. (S.Y.)

  17. Implications of crevice chemistry for cracking of BWR recirculation inlet safe-ends

    International Nuclear Information System (INIS)

    A mathematical model has been utilized to predict the electrochemistry within crevices exposed to BWR environments (0.3 μS/cm and 300oC). For both Normal Water Chemistry (NWC) and Hydrogen Water Chemistry (HWC), the modeled chemistry within the crevice was assessed to see if initiation of stress corrosion cracking was probable. (author)

  18. Compuestos orgánicos presentes en aguas de escombreras de carbón.

    OpenAIRE

    Llamas Borrajo, Juan F.; Canoira López, Laureano; García-Martínez, Mª Jesús; Ortega Romero, Marcelo; Llamas Lois, Alberto; Diaz Bautista, Maria Aranzazu

    2008-01-01

    Se ha realizado un estudio de la influencia de las aguas acidas de mina en la movilización de los hidrocarburos policidicos aromáticos (PAHs) presentes en las escombreras de carbón. Se observa como dichas aguas acidas disuelven preferentemente a los PAHs de menor peso molecular. También se ha realizado un análisis de riesgos que indica que existe un riesgo real por uso y consumo, en el interior de las viviendas, del agua subterránea. Por último, se han establecido las bases para una técnica p...

  19. Opciones de valorización de lodos de distintas estaciones depuradoras de aguas residuales

    OpenAIRE

    Colomer Mendoza, Francisco José; GALLARDO IZQUIERDO, ANTONIO; Fabián ROBLES MARTÍNEZ; Bovea Edo, María Dolores

    2010-01-01

    La legislación actual en la Unión Europea recomienda que los lodos de depuradora de aguas residuales se utilicen como fertilizante. Sin embargo, la presencia de metales pesados puede limitar esta aplicación, lo cual es más probable en caso de lodos de aguas residuales industriales. Por lo tanto, una segunda opción es la recuperación energética de los lodos. En el presente trabajo, se han analizado lodos digeridos procedentes de siete distintas depuradoras de aguas residuales españ...

  20. USO DE AGUA, POTENCIAL HÍDRICO Y RENDIMIENTO DE CHILE HABANERO (CAPSICUM CHINENSE JACQ.)

    OpenAIRE

    Wendy C. Quintal Ortiz; Alfonzo Pérez-Gutiérrez; Luis Latournerie Moreno; Cesar May-Lara; Esaú Ruiz Sánchez; Armando J. Martínez Chacón

    2012-01-01

    En las plantas el agua constituye típicamente de 80 a 95 % de la masa de los tejidos en crecimiento, donde desempeña funciones esenciales. La baja disponibilidad de agua en el suelo provoca el estrés abiótico de mayor incidencia en el crecimiento vegetal que en los sistemas agrícolas representa en pérdidas económicas. Es importante entonces estimar los requerimientos hídricos de los cultivos para mejorar su potencial productivo y el uso del agua. En este estudio se evaluó el efecto de cinco n...

  1. ECOFICCIONES E IMAGINARIOS DEL AGUA Y SU IMPORTANCIA PARA LA MEMORIA CULTURAL Y LA SOSTENIBILIDAD

    OpenAIRE

    Eloy Martos Núñez; Alberto Martos García

    2013-01-01

    La cultura del agua debe ser vinculada de forma especial a las manifestaciones del patrimonio cultural intangible de los pueblos, como las tradiciones orales o escritas, la simbología o los rituales, que conforman lo que llamamos los Imaginarios del Agua. Estos deben ser analizados y deconstruidos a la luz de los nuevos paradigmas, como la hermenéutica y la ecocrítica. De este modo, la mitografía ayuda a perfilar el significado profundo de la cultura del agua ante las nuevas demandas medioamb...

  2. Análisis comparativo de indicadores de la calidad de agua superficial

    OpenAIRE

    Jiménez Jaramillo, Mario Alberto; Vélez Otálvaro, María Victoria

    2009-01-01

    En este trabajo se hace un análisis comparativo de indicadores de calidad de agua como un acercamiento a la caracterización general de la calidad de las aguas superficiales de la cuenca de la quebrada Doña María localizada en jurisdicción del Área Metropolitana del Valle de Aburrá (AMVA) y de la Corporación Autónoma Regional del Centro de Antioquia (CORANTIOQUIA). Para ello se ha empleado el índice de calidad de agua, WQINSF, propuesto por la National Sanitation Foundation (NS...

  3. Biomateriales sorbentes para la limpieza de derrames de hidrocarburos en suelos y cuerpos de agua

    OpenAIRE

    Ortíz González, Diana Paola; Fonseca, Fabio Andrade; Rodríguez Niño, Gerardo; Montenegro Ruiz, Luis Carlos

    2010-01-01

    Este estudio esta encaminado a identificar y evaluar materiales orgánicos naturales que puedan ser utilizados como sorbentes en las operaciones de limpieza de derrames de hidrocarburos tanto en suelos como en cuerpos de agua. Se evaluó la capacidad de sorción de tres materiales: fibra de caña, fibra de coco y buchón de agua, con tres hidrocarburos: 35, 30 y 25°API, y dos tipos de agua: destilada y marina artificial, adaptando la norma ASTM F-726 y siguiendo la metodología sugerida por el pr...

  4. Durabilidad de Hormigones con adición de lodos secos de depuradora de aguas residuales

    OpenAIRE

    Albareda Parés, Francesc

    2003-01-01

    En esta tesina se estudia la influencia en la durabilidad del hormigón el hecho de incorporar al mismo, como adición, fangos secos de la depuradora de aguas residuales de Sabadell (Barcelona). Se plantea la posibilidad de utilizar los lodos como adición más en el hormigón, para conseguir inertizarlos y darles una reutilización. Para evaluar la durabilidad del hormigón se han realizado cinco ensayos: humeda-sequedad con agua potable, ataque con disolución de sulfatos, ataque con agua de mar...

  5. ELIMINACIÓN CATALÍTICA DE NITRATOS EN CONTINUO EN AGUAS NATURALES

    OpenAIRE

    Franch Martí, Cristina

    2011-01-01

    La contaminación por nitratos de las aguas naturales es un problema que afecta a gran parte de zonas agrícolas y ganaderas. Para reducir los nitratos presentes en dichas aguas se han propuesto varias técnicas, la mejor de las cuales podría ser desde un punto de vista medioambiental la reducción catalítica de nitratos a nitrógeno. En este sentido, en esta tesis se ha utilizado la técnica antes nombrada para eliminar nitratos de aguas naturales utilizando un reactor en continuo. Para ello, s...

  6. Parametrización oceanográfica en índices de calidad del agua

    OpenAIRE

    de Armas Tápanes, J.; Rodríguez Romero, R.; Zayas Suárez, D.; Fernández Vila, L.J.; López García, D.

    2010-01-01

    En el trabajo se discute la polémica situación actual de la utilización de Índices de Calidad del Agua para ecosistemas marinos, sobre la base del análisis de 20 Índices de Calidad del Agua (ICA) basados en variantes de la magnitud media, ya sea aritmética simple, ponderada o geométrica: seis norteamericanos, siete colombianos, un mexicano, cuatro holandeses y dos cubanos, y la utilización en ellos, de 31 parámetros físico-químicos (28 en agua y tres en sedimentos). Se evidenció, ...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  9. Design and axial optimization of nuclear fuel for BWR reactors

    International Nuclear Information System (INIS)

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  10. Radial optimization of a BWR fuel cell using genetic algorithms

    International Nuclear Information System (INIS)

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U235 and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix the placement of

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m3/hour with cost US$ 0.58/m3. The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  14. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  15. Reflexiones sobre la codificación del derecho de aguas

    OpenAIRE

    Mathus Escorihuela, Miguel

    2014-01-01

    El trabajo aborda la cuestión de la sistematización del derecho de aguas através de códigos de agua en la República Argentina, es decir, considerando el esquema federal de competencias del país. Trata la gestión histórica del derecho provincial de aguas y de la codificación en particular. Define que debe entenderse por un código de aguas y cuáles son las temáticas que debe contener, a la vez que se ocupa de enunciar otros temas que algunos códigos incorporan, analizando la conveniencia de su ...

  16. La ley de aguas de Entre Rios : La necesidad de su estudio en dos dimensiones

    OpenAIRE

    Maiztegui Martínez, Horacio Francisco

    2014-01-01

    El tema asignado, es el análisis de la ley de aguas de Entre Ríos N° 9.172. Trataremos de responder a las siguientes preguntas ¿Que tipos de aguas regula la ley de aguas?; ¿como se organiza el Gobierno del Agua?; ¿Como es el sistema de permisos, concesiones?; ¿Que instrumentos de conocimiento y protección prevé la ley? La ley Entrerriana, presenta una falta de adecuación a la ley Nacional como la ley 25.688, a los principios de la sustentabilidad (art.41 ° de la C.N., Ley n ° 25675 general...

  17. Claves y oportunidades para un pacto del agua en España.

    OpenAIRE

    Garrido Colmenero, Alberto; Custodio, Emilio

    2012-01-01

    Las peculiaridades geográficas españolas y su errática gestión del agua en las últimas décadas la sitúa en el marco internacional como un país incapaz de armonizar su escasez de recursos hídricos con su alto consumo interior. Por otro lado, España es pionera en tecnología relacionada con el agua, situándose entre los diez países más relevantes a nivel mundial. Por esto, diversas organizaciones independientes como el Observatorio del Agua de la Fundación Botín abogan por un Pacto del Agua, que...

  18. La biofiltración, una alternativa para la potabilización del agua

    Directory of Open Access Journals (Sweden)

    Álvaro Arango Ruiz

    2004-01-01

    Full Text Available Actualmente, la población de los países en desarrollo experimenta la carencia de condiciones sanitarias adecuadas, incluyendo la de contar con agua potable que satisfaga sus condiciones básicas. La falta de potabilización del agua está asociada con enfermedades y muerte en población infantil por enfermedades de origen hídrico. La no disponibilidad de agua apta para el consumo humano plantea la necesidad de interesarse por proyectos innovadores en este campo, como es el caso de la biofiltración, presentando alternativas de tratamiento, potabilización y distribución eficiente del agua que sean accequibles y económicamente viables

  19. Acuaporinas: proteínas mediadoras del transporte de agua.

    OpenAIRE

    Julio César Sánchez

    2009-01-01

    Introducción. Todos los organismos vivos están compuestos en su mayoría por agua, siendo ésta fundamental para la homeostasis celular a todo nivel. Por esta razón su transporte a través de las membranas biológicas ha sido siempre un campo de gran interés en la fisiología. La investigación acerca de este tópico se ha incrementado notablemente en los últimos años a partir del descubrimiento de las acuaporinas, las cuales han permitido comprender mejor los mecanismos que la célula utiliza para e...

  20. Servicios ambientales, agua y economía

    Directory of Open Access Journals (Sweden)

    Juan Pablo Ruiz Soto

    2007-11-01

    Full Text Available Los servicios ambientales (SA asociados a los ecosistemas naturales están seriamente amenazados por acciones locales y por el calentamiento global. El SA asociado al agua y su efecto sobre calidad y cantidad en cuencas hidrográficas aportantes a acueductos y sistemas de riego, son de gran importancia económica y social. Para Bogotá-Colombia el SA de los páramos y bosque nublado en la cuenca de Chingaza representa US$18,2 millones anuales. Para el abastecimiento de agua a la ciudad de Quito-Ecuador, el efecto negativo del cambio climático (CC asociado a la contracción de los glaciares representa un incremento en la inversión de un 31% versus el escenario sin CC. Se deben tomar medidas para asegurar la permanencia de los SA y mitigar los efectos del CC./ Environmental services (ES related to natural ecosystems are seriously threatened by local activities and global warming. The ES related to water and its effect on quality and quantity in watersheds that supply aqueducts and irrigation systems has great social and economic importance. For Bogotá-Colombia the ES of páramos and cloudy forest in the Chingaza basin represents US$18.2 million per year. For water supply in the city of Quito-Ecuador, the negative effect of climate change (CC related to the contraction of glaciers represents a 31% increase in investments versus the scenario without CC. Measures must be taken to ensure the permanence of ES and mitigate the effects of CC.

  1. ¿Qué tan importante es el agua? Desarrollo sustentable y recursos naturales

    OpenAIRE

    Kras, Eva; Simonsen, Ove

    2001-01-01

    A partir de una definición de desarrollo sustentable, se presenta el modelo de desarrollo global como filosofía imperante en torno al agua, y las realidades globales del agua, particularmente en torno a la contaminación, la salud, y la sustentabilidad y seguridad. Se analizan las oportunidades y retos para los países en desarrollo de América Latina, y se proponen acciones puntuales.

  2. Productividad del agua en algunos cultivos agrícolas en Cuba

    Directory of Open Access Journals (Sweden)

    Felicita González Robaina

    2014-01-01

    Full Text Available Para el mejor uso económico y social del agua se requieren de métodos para evaluar su productividad, a fin de tomar mejores decisiones en cuanto a políticas y estrategias de utilización de manera sostenible. Los administradores del agua para el riego necesitan iden- tificar tendencias en los patrones de uso y niveles de eficiencia con vistas a fijar metas y mejorar la productividad por unidad de volumen utilizado y/o consumido. El objetivo de este trabajo fue presentar los indicadores de productividad agronómica del agua de 17 cultivos a partir de la información obtenida en más de 100 experimentos de campo disponibles en la base de datos sobre necesidades hídricas del Instituto de Investigaciones de Ingeniería Agrícola. Se determinaron los rendimientos máximo y mínimo y sus respectivos valores de consumo, agua aplicada por riego y la productividad agronómica del agua aplicada para los 17 cultivos estudiados. Cultivos como los frijoles, maíz, boniato y los plátanos obtienen incrementos de rendimiento superiores al 100% cuando se aplica el agua necesaria para satisfacer su demanda hídrica, mientras que la yuca obtuvo valores menores del 50%. Los altos valores obtenidos de productividad en base al agua utilizada como riego no son un buen indicador del uso eficiente del agua por las plantas para las condiciones de la zona de estudio, donde se informan para los años analizados, porcentajes de lluvia aprovechable de hasta el 41%, supliendo ésta una parte importante de las necesidades de los cultivos agrícolas.

  3. Diseño de una planta depuradora de aguas residuales urbanas

    OpenAIRE

    Rodríguez Ruiz, Lucía

    2005-01-01

    El presente proyecto tiene por objeto el diseño de una estación depuradora de aguas residuales urbanas situada próxima a un núcleo urbano y cuyo vertido se hará en una zona considerada como zona sensible según el Real Decreto-Ley 11/1995, de 28 de diciembre, por el que se establecen las normas aplicables al tratamiento de las aguas residuales urbanas.

  4. Agua que no has de beber… déjala correr

    OpenAIRE

    Agredo Cardona, Gustavo Adolfo

    2010-01-01

    En el panorama mundial existe preocupación por la utilización o consumo del agua, sus costos (Spiller, Savedoff, y Banco Interamericano de Desarrollo, 2000), disponibilidad, aprovechamiento o pérdida de ésta. La consideración ambiental no está relacionada con la desaparición de algunas especies o por la ausencia del agua, sino porque se pone en riesgo la vida humana.

  5. Solubilidad de acetaminofén en sistemas agua-alcohol usp-jarabe usp y agua-alcohol usp-sorbitol usp

    OpenAIRE

    Bolaños, Clara Inés; Erazo, Ernesto Alonso; Luengas, Pilar; Barbosa, Helber; Martínez, Fleming

    2009-01-01

    Se determinó la solubilidad de una muestra de acetaminofén del mercado nacional a 20.0 ± 0.5 ºC en sistemas solventes ternarios agua – alcohol USP – jarabe USP – y agua – alcohol USP – sorbitol USP, como aporte a la generación sistemática de información sobre las propiedades fisicoquímicas en estudios de preformulación de formas farmacéuticas líquidas homogéneas. Los resultados muestran que las mezclas agua – alcohol USP – jarabe USP de composición en peso 30 – 60 – 10 % y a...

  6. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  7. La hidrología y su papel en ingeniería del agua

    Directory of Open Access Journals (Sweden)

    Amanda García-Marín

    2014-07-01

    Full Text Available La Hidrología es una ciencia esencial en Ingeniería del Agua, la cual abarca un amplio abanico de temas de investigación que engloban los diversos estadios del agua en el ciclo Hidrológico, tanto en atmósfera, superficie y suelo. Con motivo del relanzamiento de la revista Ingeniería del Agua se presenta un breve artículo de carácter introductorio en el que se muestran algunas de la líneas de investigación actuales en Hidrología, dedicadas a lluvia, interceptación de agua por la vegetación, sensores en Hidrología, agua subterránea, entre otras. Dicha revisión no pretende ser exhaustiva, dado el tamaño limitado de este formato de publicación, sino motivar la publicación en Ingeniería del Agua de artículos dentro de la temática Hidrología.

  8. Distribución del agua en el Perú desde una perspectiva de cuenca

    Directory of Open Access Journals (Sweden)

    Ana Sabogal

    2009-12-01

    Full Text Available El artículo plantea la importancia de integrar el concepto de cuenca a la distribución de los recursos. Se propone integrar un trabajo detallado por cuenca que involucre el cálculo del caudal de agua, la variación de la napa freática, el caudal ecológico, los volúmenes de erosión y la distribución actual del agua, para de allí plantear la distribución del agua en la cuenca. El aprovechamiento y distribución de los recursos naturales se encuentran estrechamente vinculados a la distribución del agua. Una inadecuada distribución del agua orientada solo a satisfacer el aprovechamiento de los recursos, sin tener en cuenta los ciclos ecológicos para su reposición, puede destrozar los ecosistemas existentes.El conjunto de cuencas, como unidades geográficas, deben contribuir a una gestión de recursos que permita una adecuada distribución del agua y potencie cada región, integrándola a un sistema de distribución de oportunidades, riqueza y producción en el contexto del país.

  9. El precio del agua: ¿herramienta básica para una política sostenible del agua?

    OpenAIRE

    Massarutto, Antonio

    2003-01-01

    La reciente legislación europea adopta el concepto de la recuperación completa de costes (RCC) como directriz para la tarifación del servicio de agua. En este artículo, se analiza el concepto de RCC al tiempo que se aborda la razón económica de su aplicación desde diferentes perspectivas. La idea básica aquí desarrollada es que se trata de un concepto un tanto confuso, ya que los costes se recuperan por definición, estando por tanto el verdadero problema en evaluar cómose recuperan. Después d...

  10. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  11. Investigación técnica y económica sobre desinfección de aguas residuales por sistemas de oxidación

    OpenAIRE

    Gordillo de Coss, Gerardo Enrique

    2013-01-01

    Un incremento de la demanda del agua, junto con el aumento de la contaminación, ha provocado que hoy en día la reutilización de las aguas depuradas sea necesaria, pero la reutilización de aguas debe garantizar y minimizar los posibles riesgos sanitarios y medioambientales que su práctica pueda provocar. En España estos parámetros se encuentran regulados por el RD 1620/2007 relativo al régimen jurídico de la reutilización de las aguas depuradas. Las aguas regeneradas son aguas que ya han sido ...

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

    International Nuclear Information System (INIS)

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

  13. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    Science.gov (United States)

    Trianti, Nuri; Su'ud, Zaki; Arif, Idam; Riyana, EkaSapta

    2014-09-01

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  14. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    International Nuclear Information System (INIS)

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  16. Estudio del mecanismo de falla de terraplenes debido a la infiltración de aguas lluvias mediante el monitoreo de presiones de poros y contenidos de agua

    OpenAIRE

    García, Edwin; Uchimura, Taro

    2008-01-01

    El presente artículo muestra experimentos realizados utilizando modelos a escala para estudiar el proceso de infiltración en terraplenes sometidos a aguas lluvias. Varios modelos a escala fueron construidos usando un suelo arenoso con alto contenido de finos; para aplicar la lluvia artificial y observar el proceso de infiltración del agua se utilizó un tubo de irrigación. Para monitorear los cambios en succión ocasionados por los ciclos de humedecimiento y secado se instalaron dentro del suel...

  17. Estudio del mecanismo de falla de terraplenes debido a la infiltración de aguas lluvias mediante el monitoreo de presiones de poros y contenidos de agua

    OpenAIRE

    EDWIN GARCÍA; TARO UCHIMURA

    2007-01-01

    El presente artículo muestra experimentos realizados utilizando modelos a escala para estudiar el proceso de infiltración en terraplenes sometidos a aguas lluvias. Varios modelos a escala fueron construidos usando un suelo arenoso con alto contenido de finos; para aplicar la lluvia artificial y observar el proceso de infiltración del agua se utilizó un tubo de irrigación. Para monitorear los cambios en succión ocasionados por los ciclos de humedecimiento y secado se instalaron den...

  18. Estudio geoquímico de suelos y aguas como base para evaluar la contaminación: relación roca-suelo-agua

    OpenAIRE

    Vázquez Garranzo, Icíar

    2008-01-01

    Los objetivos generales del estudio de los suelos han sido, por un lado, establecer los valores de referencia de elementos traza a partir de la determinación de los valores de fondo y, por otro, demostrar las diferencias de disponibilidad entre el contenido de un elemento aportado al suelo y el contenido litogénico del mismo en suelo. Para el estudio de las aguas, los objetivos han sido: establecer la calidad química del agua en base a su comparación con indicadores de acuerdo con los req...

  19. Pobreza, agua y tierra en Ambato, Ecuador: Perfil de pobreza y el acceso y manejo de agua y tierra en la cuenca de Ambato, Ecuador

    OpenAIRE

    Chapalbay, Washington; Cordero, Doris; Gómez, Ligia Ivette; Munk Ravnborg, Helle

    2007-01-01

    Este informe tiene coma objetivo presentar el perfil de pobreza para la cuenca del rio Ambato, la prov1ncia de Tungurahua, Ecuador, )' analizar en que medida el acceso al agua, particularmente para el riego, ya la tierra, depende de los niveles de pobreza. El manejo del agua es clave en la cuenca de Ambato, principalmente para la producci6n agricola. Con base en el perEl de pobreza, casi dos terceras partes de las familias tienen acceso al riego. De la misma forma, casi la totalidad de las fa...

  20. La posible creación de mercados de agua y la gobernabilidad de este recurso en Colombia. Comentarios sobre el proyecto de ley del agua

    OpenAIRE

    Manuel Rodríguez Becerra

    2005-01-01

    En la actualidad se está tramitando un proyecto de Ley del Agua en el Congreso de la República, presentado a su consideración por el Gobierno Nacional, mediante el cual "se establecen medidas para orientar la planificación y administración del recurso hídrico en el territorio nacional." Se trata de la segunda reforma del gobierno del Presidente Alvaro Uribe relacionada con la gestión del agua en Colombia. En efecto, al principio de su gobierno se expidió la Ley de 2003 de creación del Ministe...

  1. EFECTOS DE LA PRIVACIÓN DE AGUA EN UN PROGRAMA DE REFORZAMIENTO IV 5' SOBRE EL PESO CORPORAL, EL CONSUMO DE AGUA Y ALIMENTO EN RATAS

    OpenAIRE

    Antonio López-Espinoza; Américo Ríos; Ma. Eugenia Soto

    2004-01-01

    Cuatro ratas fueron expuestas a un programa de privación de agua con 1 hora de acceso libre y 23 horas de privación en un programa de reforzamiento N 5; otras cuatro ratas fueron usadas como sujetos control. Laduración de los periodos de restricción fuede 30, 64, 25 Y 28 días con intervalos de libre acceso entre cadaunode ellos. El alimento se mantuvo en libreacceso durante todoel experimento. Los resultaron confirmaron que al retirar la privación de agua se modificó el patrón de alimentación...

  2. Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

    OpenAIRE

    2008-01-01

    Boiling water reactor (BWR) instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presen...

  3. Dispositivo electrónico de medición del caudal de agua para canales abiertos

    OpenAIRE

    Julio Cesar Ayala López; Roberto Albóniga Gil

    2015-01-01

    El método de riego por inundación, que desplaza el agua mediante canales abiertos y surcos ha sido usado durante años en la producción agrícola, aunque requiere una gran cantidad de agua y su eficacia no es muy alta ya que la mayoría del agua no se puede extraer directamente en las raíces de las plantas. Siendo de vital importancia lograr datos de campo confiables y lo suficientemente precisos que nos permitan estudiar y proyectar manejos del agua con el menor grado de incertidumbre posi...

  4. Normalización de la terminología mexicana sobre el agua Standardization of Mexican terminology on water

    OpenAIRE

    Catalina Naumis-Peña; Verónica Vargas-Suárez

    2009-01-01

    Se expone el resultado de una investigación llevada acabo para conocer el estado de la normalización terminológica en tesauros especializados que incluyen el agua dentro de sus áreas contratando los términos usados en la literatura mexicana. La documentación que se genera sobre el tema del agua en México crece en forma exponencial en tanto que la problemática del agua exige soluciones integrales. Actualmente no existe un tesauro, ni terminologías relacionadas con el tema del agua en México, n...

  5. La lucha por el agua en el país de la lluvia (Galicia, siglos XVI-XIX)

    OpenAIRE

    Ofelia REY CASTELAO

    2013-01-01

    En este artículo se estudia la intensa conflictividad por el agua que se vivió en la Galicia de la Edad Moderna debido al déficit de aguas de riego durante el verano, y a los complejos sistemas de usufructo del agua para uso doméstico y agrícola. Para conocer las causas de los conflictos, su cronología, su distribución territorial y los grupos sociales que los protagonizaron, se utilizan escrituras notariales –sobre todo, poderes y concordias–, y los pleitos por aguas atendidos por la Real Au...

  6. Casos de estudio: ahorro de agua en los regadíos de Aragón

    OpenAIRE

    Playán Jubillar, Enrique; Faci González, José María; Cavero Campo, José; Dechmi, Farida; Lecina Brau, Sergio

    2000-01-01

    En esta ponencia se comienza discutiendo las posibilidades de ahorro de agua en la agricultura. Se concluye que sólo es posible ahorrar agua si se disminuye la intensidad de los cultivos o su extensión. Por ello, se opta por la "conservación" de agua, un concepto que implica el ahorro de agua regulada. Se comentan cinco casos de estudio de zonas regables en Aragón. En Almudévar (un regadío tradicional), la eficiencia de riego está en torno al 50%, y las posibilidades de conserv...

  7. Calidad de las aguas meteóricas en la ciudad de Itajubá, Minas Gerais, Brasil

    OpenAIRE

    Herlane Costa Calheiros; Mônica Regina Gomes; Percy Máximo Anco Estrella

    2014-01-01

    El objetivo de este estudio fue evaluar la calidad del agua meteórica en la ciudad de Itajubá, ubicado en la parte sur de Minas Gerais, Brasil. Para ello, se realizaron la colecta y análisis de la calidad de las muestras de agua de lluvia después de escurrir sobre la superficie de tres tipos de materiales; amianto, zinc y cerámica. También se realizó una colecta directa para comparar los resultados y verificar los cambios en la calidad del agua. Se observó que la calidad del agua de lluvia va...

  8. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de inyeccion de agua de refrigeracion a baja presion (LPCI) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C., E-mail: renedelgado2015@hotmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

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

    International Nuclear Information System (INIS)

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

  10. Investigation of distorted geometry simulation of pool dynamics in horizontal-vent BWR containments

    International Nuclear Information System (INIS)

    Experiments were conducted to investigate the accuracy of distorted geometry testing of pool dynamics in horizontal-vent BWR containments. Distorted-geometry testing implies testing in systems where the flow-wise dimensions are full scale, but all dimensions transverse to the flow are reduced in the same proportion. The assumption is that flow velocities, pressures and other thermodynamic properties will be the same in the distorted-geometry system as in its correctly proportioned counterpart. The experiments, which were done at small scale using the established scaling laws, showed that the geometric distortions can have a significant effect on the pool swell under conditions which are roughly representative of horizontal-vent BWR containment systems during a LOCA. Breakthrough occurred later, the water ligament was thicker, and pool velocity lower in a system where the cross-sectional areas were reduced by a factor of three. Some reasons for the differences are discussed

  11. Aggressive chemical decontamination tests on small valves from the Garigliano BWR

    International Nuclear Information System (INIS)

    In order to check the effectiveness of direct chemical decontamination on small and complex components, usually considered for storage without decontamination because of the small amount, some tests were performed on the DECO experimental loop. Four small stainless steel valves from the primary system of the Garigliano BWR were decontaminated using mainly aggressive chemicals such as HC1, HF, HNO3 and their mixtures. On two valves, before the treatment with aggressive chemicals, a step with soft chemical (oxalic and citric acid mixture) was performed in order to see whether a softening action enhances the following aggressive decontamination. Moreover, in order to increase as much as possible the decontamination effectiveness, a decontamination process using ultrasounds jointly with aggressive chemicals was investigated. After an intensive laboratory testing programme, two smaller stainless steel valves from the primary system of the Garigliano BWR were decontaminated using ultrasounds in aggressive chemical solutions

  12. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  13. Discussion on 'Electrochemical potential measurements under simulated BWR water chemistry conditions'

    International Nuclear Information System (INIS)

    In the above-referenced paper, Lin et al. report measurements of the corrosion potentials (the electrochemical potential or ECP) of types 304 and 316 SS in simulated boiling water reactor (BWR) heat transport environments at 270 C. There are four reasons for this discussion: to demonstrate that their theoretical explanation for the variation of ECP with oxygen concentration is inadequate; to show that their flow velocity/ECP results for oxygenated and hydrogenated systems are experimentally inconclusive because of experimental problems and, in any case, are inconsistent with electrochemical expectations; to cite previous work on the origin of the ECP of stainless steels in BWR environments that was not referenced in the paper but provides a basis for interpreting their data; and to identify previous work on the effect of Cu2+ on the ECP of type 304 SS, which was also not referenced in the paper

  14. Developing a basis for predicting and assessing trends in BWR core tracking

    International Nuclear Information System (INIS)

    In the process of designing a boiling water reactor (BWR) fuel cycle, it is necessary to estimate the bias eigenvalue trend or nuclear design basis, which has a large effect on the cycle parameters. Currently, the history of previous cycles is used as a basis, but due to increasing demands of higher energy output per cycle, and unexpected events during a cycle, predictions become challenging. To aid in developing a basis for making predictions, various perturbations in the areas of fuel manufacturing and plant measurement were studied in a multicycle analysis. These perturbations have a range of effects on several different cycle parameters. The results of this study are intended to assist in the prediction and assessment of trends in the BWR industry. (authors)

  15. Development of membrane moisture separator for BWR off-gas system

    International Nuclear Information System (INIS)

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

  16. Crack growth behavior of irradiated austenitic stainless steels in BWR environments

    International Nuclear Information System (INIS)

    Crack growth tests have been performed in boiling water reactor (BWR) environments on Types 304 and 316 stainless steel that were irradiated to fluence levels up to 2.0 x 1021 n cm-2 (E > 1 MeV) at approx. 288 degC in a helium environment. Two waveforms were used in the tests, slow/fast sawtooth and trapezoidal. The cyclic loading was done with rise times between 30 and 1000 s. At the longer rise times, the environmental contributions to the crack growth rate dominate. The trapezoidal waveform essentially represents constant load with periodic unloading and loading. The results indicate significant enhancement of crack growth rates of the irradiated steel in the BWR environment with normal water chemistry. The effects of fluence and hydrogen water chemistry are presented. (author)

  17. Assembly Based Modular Ray Tracing and CMFD Acceleration for BWR Cores with Different Fuel Lattices

    International Nuclear Information System (INIS)

    The geometry module of the DeCART direct whole core calculation code has been extended in order to analyze BWR cores which might have a mixed loading of different fuel types. First, an assembly based modular ray tracing scheme was implemented for the Method of Characteristic (MOC) calculation, and a CMFD formulation applicable for unaligned mesh conditions was then developed for acceleration the MOC calculation. The new calculation feature has been validated by comparing DeCART BWR assembly calculations with the MCU Monte Carlo calculations. A good agreement identified by the maximum eigenvalue difference of 120 pcm and the maximum pin power error of about 1% has been achieved. The CMFD scheme is shown to reduce the number of MOC iterations by factors of 12-25 without loss of accuracy. (authors)

  18. BWR 90: An evolutionary ABWR plant for the next decade(s)

    International Nuclear Information System (INIS)

    The future of nuclear power is determined by ''conflicting'' considerations. It must be generated at acceptable cost with negligible environmental effects, and it must achieve and maintain a reasonable level of public acceptance. The latter is, however, not necessarily governed by rational assessments. The ABB approach to ''nuclear'' can be characterized as a ''cautious evolution''. In the next decade(s), its offerings will largely be based on ''evolutions'' of two successful LWR types, the BWR 90 and System 80+. Both designs can be designed, licensed and constructed in accordance with any safety regulations in force or envisaged in the Western world. This paper elaborates on BWR 90, a design that is closely based on its forerunner, taking into account experiences from design and engineering, construction and commissioning, and operation, needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings

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

    International Nuclear Information System (INIS)

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

  20. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH Univ. of Applied Sciences, Deggendorf (Germany)

    2014-07-01

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation programme was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment with integrated pressure suppression system. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The main target was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. (orig.)

  1. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    International Nuclear Information System (INIS)

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation programme was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment with integrated pressure suppression system. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The main target was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. (orig.)

  2. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  3. Droplet entrainment and deposition rate models for determination of boiling transition in BWR fuel assembly

    International Nuclear Information System (INIS)

    Droplet entrainment and deposition rates are of vital importance for mechanistic determination of critical power and location of boiling transition in a BWR fuel assembly. Data from high-pressure, high-temperature steam-water adiabatic experiments conducted in very tall test sections are used to develop a combination of equilibrium entrainment-deposition rate. Application of this combination to the heated tests conducted in a shorter test section of typical height of a BWR fuel assembly shows that correct split of total liquid in form of the film and droplets at the onset of annular-mist flow regime is also important to obtain good prediction of film flow rates/entrainment fraction. The improved model is then applied to simulate critical power tests in annulus and rod bundles. (author)

  4. Applied methods for mitigation of damage by stress corrosion in BWR type reactors

    International Nuclear Information System (INIS)

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

  5. On-line critical control rod pattern prediction algorithm for BWR plant startup

    International Nuclear Information System (INIS)

    This paper describes an on-line algorithm for predicting the critical control rod pattern, which has been developed to reduce the mental strain on operators while withdrawing control rods in the BWR plant startup operation. The proposed algorithm estimates a target eigenvalue (eigenvalue bias) for a three-dimensional neutron kinetics model with a neutron source incorporating actual neutron detector readings. The critical control rod pattern is then predicted based on the estimated eigenvalue bias. The algorithm has been verified using data obtained from an actual startup operation on a BWR model-5 plant, and the estimated eigenvalue bias agreed well with the effective multiplication factor at the criticality actually determined from the operator's judgement. (author)

  6. The stability analysis of LAPUR6 for Chinshan BWR nuclear power plant

    International Nuclear Information System (INIS)

    Chinshan Nuclear Power Plant (NPP) is the first NPP in Taiwan which is a BWR/4 plant. The original rated power for each unit was 1775 MWt. After the project of measurement uncertainty recovery (MUR) for Chinshan NPP, the operating power is 1805 MWt now. The Chinshan NPP Unit 2 cycle 23 stability analyses were performed by the LAPUR6 stability analysis methodology. Comparing the LAPUR6 stability analysis results and vendor's results, they are similar. (author)

  7. A Deterministic/probalistic analysis of Ex-Vessel melt risk in a BWR

    OpenAIRE

    Abal López, Javier

    2006-01-01

    The present study is concerned with deterministic and probabilistic analysis of ex-vessel melt risks in a Swedish designed BWR plant. The focus is placed on a station blackout (SBO) scenario, with immediate SCRAM and subsequent activation of the main steam valve isolation (at 52 s). Four sequences were examined in detail to study the effect of two valves systems related to the operation of ADS (Automatic Depressurization System), and cavity flooding by water from suppression po...

  8. Transient boiling and void formation during postulated reactivity-initiated accident in BWR: Experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (author)

  9. BWR-X experimental reactor for operating in remote sites (thermal-hydraulic study)

    International Nuclear Information System (INIS)

    The primary objective of this indigenous nuclear plant is to power supply no ease accessible places needed of support, like scientific bases and government facilities. What's more, design and built this particular type BWR at low pressure and temperature will give substantial background on nuclear engineering. The paper is intended to treat the nuclear reactor as a heat source in a power cycle and to study the processes by which energy is removed and converted to useful power. (author)

  10. SIMULATE-3K simulation of the Ringhals 1 BWR stability measurements

    International Nuclear Information System (INIS)

    SIMULATE-3K is the transient analysis version of the SIMULATE-3 advanced nodal reactor analysis code. The transient form of the 3-D QPANDA nodal neutronics model has been coupled to a 3-D channel thermal-hydraulics model and a 1-D transient excore peripheral systems model. This paper presents comparisons of SIMULATE-3K calculations and measured BWR stability data from Ringhals Unit 1, Cycles 14-17, as formulated by the OECD/NEACRP. (author)

  11. Proving test on thermal-hydraulic performance of BWR fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8 x 8, 9 x 9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPEC-TH-B Project). The high-burnup 8 x 8 fuel (average fuel assembly discharge burnup: about 39.5 GWd/t), has been utilized from 1991. And the 9 x 9 fuel (average fuel assembly discharge burnup: about 45 GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9 x 9 fuel assembly. Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9 x 9 fuel combined with the previously reported results of high-burnup 8 x 8 fuel. As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed. (author)

  12. Transient boiling and void formation during postulated reactivity initiated accident in BWR: experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (authors)

  13. Propagation of cracks by stress corrosion in conditions of BWR type reactor

    International Nuclear Information System (INIS)

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  14. Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

    International Nuclear Information System (INIS)

    As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

  15. BWR Mark I pressure suppression study: effect of downcomer fill level on the vertical load function

    International Nuclear Information System (INIS)

    The investigation reported forms a part of the BWR Mark I Pressure Suppression Experiment Program and is one of a series of small scale studies designed to evaluate limited aspects of the pool dynamics phenomena prior to conduct of the 1/5 scale air test series. Presented is an experimental study of the effect of downcomer fill level (DFL) on the vertical load function

  16. Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, A.; Sanchez, V. [Karlsruhe Inst. of Technology, Inst. for Neutron Physics and Reactor Technology, Herman-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Hoogenboom, J. E. [Delft Univ. of Technology, Faculty of Applied Sciences, Mekelweg 15, 2629 JB Delft (Netherlands)

    2012-07-01

    As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

  17. Role of BWR secondary containments in severe accident mitigation: issues and insights from recent analyses

    International Nuclear Information System (INIS)

    All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by a secondary containment consisting of a reactor building and refueling bay (MK I and MK II designs), a shield building, auxiliary building and fuel building (MK III), or an auxiliary building and enclosure building (Grand Gulf style MK III). Although secondary containment designs are highly plant specific, their purpose is to minimize the ground level release of radioactive material for a spectrum of traditional design basis accidents. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident. This paper presents a brief overview of domestic BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent ORNL secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented

  18. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    International Nuclear Information System (INIS)

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  19. Application of a lattice filter algorithm to BWR noise monitoring and diagnosis

    International Nuclear Information System (INIS)

    a method for BWR noise monitoring and diagnosis based on an adaptive lattice filter is presented in this paper. It is specifically intended for applying to on-line real time BWR stability monitoring during reactor operation relatively close to unstable regimes, where fast tracking of any stability state change is essential. The authors present a set of algorithms for estimating stability parameters, i.e. decay ratio and peak frequency, for detecting anomaly occurrences based on the pattern recognition methodology, and for calculating signal DC component and mean amplitude. A unique feature of the method is the unified approach for the signal processing, in that these algorithms are all executed based on model identification using the adaptive lattice filter. The present method was applied to BWR noise data collected during stability tests at Forsmark 1 and 2 in Sweden. The result demonstrates that the present method is capable of fast tracking of stability state changes. The paper stresses that the evaluation of the process state by combining the results from the stability parameters estimation, pattern comparison, and checking the behavior of DC component and mean amplitude of process signals is important for obtaining reliable estimate and adequate information on the stability state, especially when the reactor is operated near instability

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

    International Nuclear Information System (INIS)

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

  1. Assessment of the interchannel mixing model with a subchannel analysis code for BWR and PWR conditions

    International Nuclear Information System (INIS)

    The influence of the interchannel mixing model employed in a traditional subchannel analysis code was investigated in this study, specifically on the analysis of the enthalpy distribution and critical heat flux (CHF) in rod bundles in BWR and PWR conditions. The equal-volume-exchange turbulent mixing and void drift model (EVVD) was embodied to the COBRA-IV-I code. An optimized model of the void drift coefficient has been devised in this study as the result of the assessment with the two-phase flow distribution data for the general electric (GE) 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margins of typical light water nuclear reactor (LWR) cores were evaluated by considering the influence on the local parameter CHF correlation and the hot channel analysis result. It appeared that the interchannel mixing model has an important effect upon the analysis of CHFR margin for BWR conditions. (orig.)

  2. Fabrication and operating history considerations in assessing relative SCC susceptibility of BWR components

    International Nuclear Information System (INIS)

    The spectrum in material, stress and environmental conditions present in BWR components results in a varied intergranular stress corrosion cracking (IGSCC) response. Fabrication techniques can significantly influence susceptibility to cracking. Components fabricated using cold forming processes tend to experience SCC initiation earlier than similar components subjected to annealing heat treatments, with the degree of acceleration related to the severity and extent of cold work. BWR field experience has shown that SCC initiation and growth was significantly accelerated by the presence of cold work in the upper steam dryer support ring components. Similarly, the propensity for IGSCC initiation and growth in Alloy 182 weld joints may be influenced by microfissuring formed during fabrication and related welding. Ultrasonic inspection results show apparent evidence of accelerated IGSCC initiation due to pre-existing microfissures in nozzle welds and butters. IGSCC is also influenced by irradiation and water chemistry conditions on a creviced component. In addition to demonstrated effects of coolant conductivity and accrued fluence on IGSCC behavior, the localized flux level apparently influences the potential for IGSCC in some components. BWR field experience shows evidence that the propensity for IGSCC in creviced and irradiated control blades is dependent on flux level in higher conductivity plants. Operating with higher purity coolant apparently significantly reduces the potential for IGSCC in creviced components exposed to high flux regions

  3. Effect of bundle size on BWR fuel bundle critical power performance

    International Nuclear Information System (INIS)

    Effect of the bundle size on the BWR fuel bundle critical power performance was studied. For this purpose, critical power tests were conducted with both 6 x 6 (36 heater rods) and 12 x 12 (144 heater rods) size bundles in the GE ATLAS heat transfer test facility located in San Jose, California. All the bundle geometries such as rod diameter, rod pitch and rod space design are the same except size of flow channel. Two types of critical power tests were performed. One is the critical power test with uniform local peaking pattern for direct comparison of the small and large bundle critical power. Other is the critical power test for lattice positions in the bundle. In this test, power of a group of four rods (2 x 2 array) in a lattice region was peaked higher to probe the critical power of that lattice position in the bundle. In addition, the test data were compared to the COBRAG calculations. COBRAG is a detailed subchannel analysis code for BWR fuel bundle developed by GE Nuclear Energy. Based on these comparisons the subchannel model was refined to accurately predict the data obtained in this test program, thus validating the code capability of handling the effects of bundle size on bundle critical power for use in the study of the thermal hydraulic performance of the future advance BWR fuel bundle design. The author describes the experimental portion of the study program

  4. Supercell burnup model for the physics design of BWR fuel assemblies

    International Nuclear Information System (INIS)

    A code called SUPERB has been developed for the BWR fuel assembly burnup analyses using supercell model. Each of the characteristic heterogeneities of a BWR fuel assembly like water gap, poisoned pins, control blade etc., is treated by invoking appropriate supercell concept. The burnup model of SUPERB is so devised as to strike a balance between accuracy and speed. This is achieved by building isotopic densities in each fuel pin separately while the depletion equations are solved only in a few groups of pins or burnup zones and the multigroup neutron spectra are differentiated in fewer group of pincell types. Multiple fuel ring burnup is considered only for Gd isotopes. A special empirical formula allows the microscopic cross section of Gd isotopes to be varied even during burnup integration. The supercell model has been tested against Monte Carlo results for the fresh cold clean Tarapur fuel assembly with two Gd fuel pins. The burnup model of SUPERB has been validated against one of the most sophisticated codes LWR-WIMS for a benchmark problem involving all the complexities of a BWR fuel assembly. The agreement of SUPERB results with both Monte Carlo and LWR-WIMS results is found to be excellent. (auth.)

  5. Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark

    International Nuclear Information System (INIS)

    Highlights: • The reference solution for Oskarshamn-2 BWR stability benchmark. • Description of TRACE/PARCS modeling of Oskarshamn-2. • Description of FW temperature correction and its effect on the calculated results. • Description of space-time convergence discretization and its effect on the calculated results. - Abstract: On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations, and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal–hydraulics (TH), and TH/NK coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one

  6. Thermal hydraulic test apparatus to develop advanced BWR fuel bundles with spectral shift rods (SSR)

    International Nuclear Information System (INIS)

    An advanced water rod (WR) called the spectral shift rod (SSR), which replaces a conventional WR in a BWR fuel bundle, enhances the BWR's merit of uranium saving through the spectral shift operation. The SSR consists of an inlet hole, a wide ascending path, a narrow descending path and an outlet hole. The inlet hole locates below a lower tie plate (LTP) and the outlet hole is set above it. In the SSR, water boils by neutron and gamma-ray heating and water level is formed in the ascending path. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. Steady state and transient tests were conducted to evaluate SSR thermal-hydraulic characteristics under BWR operation condition. The several types of SSR configuration were tested, which covers SSR design in both next generation and conventional BWRs. In this paper, the test apparatus overview and measurement systems especially two phase water level measures in the SSR are presented. (author)

  7. Effect of thermal-hydraulic feedback on the BWR rod drop accident

    International Nuclear Information System (INIS)

    An important design-basis accident for boiling water reactors (BWR's) is the rod drop accident (RDA). This accident is defined to be a rapid reactor transient caused by an accidental drop (out of the core) of the highest-worth control rod at various conditions ranging from cold start-up to about 10% of rated power. For most BWR designs the highest worth rod is normally situated at the center of the core. Despite the fact that the chance of a RDA in extremely unlikely, the consequence of the RDA is of concern because of the potential for damage to fuel rods. Neglecting moderator feedback during the RDA is a poor assumption because energy is deposited in the fuel over a 3 to 4 second time period and hence there is time for heat to be conducted to the coolant. This may tend to ameliorate the accident considerably. Evaluation of the thermal-hydraulic feedback effect on the RDS in a BWR has been scarce in the literature. The object of this paper is to demonstrate the beneficial effect of thermal-hydraulic feedback in the RDA

  8. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  9. Assessment of severe accident prevention and mitigation features: BWR, Mark III containment design

    International Nuclear Information System (INIS)

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark III containments (BWR Mark II's), have been identified. These features and actions were developed specifically for the Grand Gulf Plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark III to severe accident containment loads were also identified. In addition, those features of a BWR Mark III, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Grand Gulf and other Mark III plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

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

  13. Aging and service wear of control rod drive mechanisms for BWR nuclear plants

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''managing'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to normal wear and aging are the Graphiter seals. The predominant causes of aging for these seals are mechanical wear and thermally induced embrittlement More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are valve seals, discs, seats, stems, packing, and diaphragms. Since CRDM changeout and rebuilding is one of the highest dose, most physically challenging, and complicated maintenance activities routinely accomplished by BWR utilities, this report also highlights recent innovations in CRDM handling equipment and rebuilding tools that have resulted in significant dose reductions to the maintenance crews using them

  14. Analysis CFD for the hydrogen transport in the primary containment of a BWR

    International Nuclear Information System (INIS)

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

  15. Ahorro de agua y materia prima en los procesos de pelambre y curtido del cuero mediante precipitación y recirculación de aguas

    Directory of Open Access Journals (Sweden)

    SANDRA AGUDELO A.

    2007-01-01

    Full Text Available El principio del proceso de recuperación de cromo, consiste en precipitar el cromo contenido en el licor residual como hidróxido utilizando un álcali; este precipitado se separa por sedimentación y se disuelve nuevamente con ácido sulfúrico. De los resultados obtenidos, se observaron porcentajes de remoción superiores al 99%. Con la recirculación directa de aguas de pelambre fue posible obtener ahorros de agua del 64% y ahorros en materia prima del 32 %, además, con una adecuada selección del tamizador es posible recircular el agua un número máximo de tres veces, sin afectar la calidad del producto final. Así mismo, se comprobó la viabilidad de la recirculación directa de aguas de curtido, con ahorros de materia prima del 17%. Con medidas de este tipo, prácticas y fáciles de implementar, la industria del curtido del cuero podría disminuir considerablemente la contaminación, operar de una forma mas segura y con mayor eficiencia de producción.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  18. CFD predictions of standby liquid control system mixing in lower plenum of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, Christopher, E-mail: christopher.boyd@nrc.gov; Skarda, Raymond, E-mail: Raymond.skarda@nrc.gov

    2014-11-15

    Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

  19. Evaluación económica de la reutilización de agua para la creación de humedales artificales

    OpenAIRE

    Varela Bernaus, Héctor

    2007-01-01

    Orígenes y tratamiento de las aguas residuales. Funcionamiento de las estaciones depuradoras de aguas residuales (EDAR'S) y de los sistemas de humedales construidos (SHC'S). Evaluación económica de los (PMV) SHC --> Ejemplo: Can Cabanyes. Precio mínimo de venta del agua regenerada en los SHC y beneficios generados por los SHC. Ingresos generados por los SHC y factores limitantes del PMV y del agua regenerada.

  20. Analisis del efecto de los mercados de agua sobre el beneficio de las explotaciones, la contaminacion por nitratos y el empleo eventual agrario

    OpenAIRE

    Calatrava-Leyva, Javier; Colmenero, Alberto Garrido

    2001-01-01

    RESUMEN: La evidencia disponible sobre los efectos externos de los mercados de agua actualmente en funcionamiento es contradictoria. Algunos trabajos demuestran que el intercambio de agua reduce la contaminación derivada de los sistemas productivos del regadío, mientras que otros concluyen lo contrario. El uso combinado de tasas sobre el input contaminante y la implantación de un mercado de agua plantea incentivos a los productores con resultados ambiguos, entre otras razones porque agua y fe...