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Sample records for bwr reactors tendencias

  1. 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)

  2. 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.)

  3. 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.)

  4. 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.)

  5. 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.)

  6. 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)

  7. 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)

  8. 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)

  9. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  10. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  20. 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)

  1. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  2. The design and use of proficiency based BWR reactor maintenance and refuelling training mockups

    International Nuclear Information System (INIS)

    The purpose of this paper is to describe the ABB experience with the design and use of boiling water reactor training facilities. The training programs were developed and implemented in cooperation with the nuclear utilities. ABB operates two facilities, the ABB ATOM Light Water Reactor Service Center located in Vasteras, Sweden, and the ABB Combustion Engineering Nuclear Operations BWR Training Center located in Chattanooga, Tennessee, USA. The focus of the training centers are reactor maintenance and refueling activities plus the capability to develop and qualify tools, procedures and repair techniques

  3. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    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)

  4. Analysis of assemblies exchange in the core of a reactor BWR; Analisis del intercambio de ensambles en el nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kauil U, J. S. [Universidad Autonoma de Yucatan, Facultad de Ingenieria, Av. Industrias no contaminantes por Anillo Periferico Norte s/n, Apdo. Postal 150 Cordemex, Merida, Yucatan (Mexico); Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia del Cueto, R., E-mail: san_dino@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The performance of the core of a boiling water reactor (BWR) was evaluated when two assemblies are exchanged during the fuel reload in erroneous way. All with the purpose of analyzing the value of the neutrons effective multiplication factor and the thermal limits for an exchange of assemblies. In their realization the mentioned study was based in a transition cycle of the Unit 1 of the nuclear power plant of Laguna Verde. The obtained results demonstrate that when carrying out an exchange between two fuel assemblies in erroneous way, with regard to the original reload, the changes in the neutrons effective multiplication factor do not present a serious problem, unless the exchange has been carried out among a very burnt assembly with one fresh, where this last is taken to the periphery. (Author)

  5. Reactor scram experience for shutdown system reliability analysis. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.

    1976-06-01

    Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.

  6. Thermal hydraulics characterization of the core and the reactor vessel type BWR

    International Nuclear Information System (INIS)

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  7. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

    International Nuclear Information System (INIS)

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  8. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  9. 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)

  10. Gas/liquid separator for BWR type reactor

    International Nuclear Information System (INIS)

    A two phase gas/liquid flow generated at a heating portion of a nuclear reactor is swirled by inlet vanes. The phase gas/liquid flow uprises as a vortex flow in a vortex cylinder, and a liquid phase of a high density gathers at the outer circumference of the vortex cylinder. The liquid phase gathered at the outer circumference is collected at the inlet of a discharge flow channel which protrude into the vortex cylinder and in a three-step structure, and introduced into a recycling liquid phase passing through the discharge flow channel for liquid phase. There is provided a structure that separated liquid collected at the lowermost state in the inlet of the three-step discharge flow channel inlet descends in the discharge flow channel, then uprises in an uprising flow channel and is introduced into the recycling liquid phase by way of a discharge flow channel exit. The height of the discharge flow channel exit is determined equal to that of a liquid level of the recycling liquid phase during rated operation of the reactor. Accordingly, even in a case where the liquid level in the recycling liquid phase is lowered, the liquid level of the uprising flow channel is kept equal to that during rated operation. (I.N.)

  11. Prediction of the stability of BWR reactors during the start-up process; Prediccion de la estabilidad de reactores BWR durante el proceso de arranque

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R. [ININ, Km. 36.5 Carretera Mexico-Toluca, 52045 Salazar, Estado de Mexico (Mexico); Blazquez M, J.B. [Centro de Investigaciones Energetics, Medioambientales y Tecnologicas, Av Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  12. Identification of chromium oxides and other solids in BWR reactor water

    International Nuclear Information System (INIS)

    Radioactive solid particles in reactor water may deposit as hot spots on reactor component surfaces, contributing to plant radiation field build-up. Phase identification of these solid particles would improve our understanding about the origins of the 'hot spots' and their behaviour under various water chemistry conditions. Phase identification is also important for the purpose of experimental verification of some thermodynamic calculations that predict thermodynamic stability of certain solid phases in BWR water environments. This paper concerns a transmission electron microscopy study on solid particles that were collected from two Swedish BWRs operated with hydrogen water chemistry. In the samples collected from both reactors, a significant fraction of the total activities came from radionuclide Cr-51. Among various solid particles detected, a significant number of chromium oxide particles were found. From one reactor amorphous chromium oxide particles were detected while from another reactor crystalline Cr2O3 was found. The presence of the metastable amorphous chromium oxide in the coolant suggests that any assumption of achieving thermodynamic equilibrium in the coolant system would not be valid. (author)

  13. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    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)

  14. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  15. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    for a large reactor could reach 20 gigabytes) that it is not possible to load into RAM memory of an operating system with 32 bit architecture. A special procedure has been developed within the MATLAB environment to remove this memory limitation, and to invert such large matrices and finally obtain the reactor transfer functions that enable the study of system stability. Various applications of the present frequency-domain code to a typical BWR fuel assembly, a BWR core, and to a chemical reactor showed a good agreement with reference results. (author)

  16. Evaluation of pressure transitories in BWR type reactors using the BWRDYN code; Evaluacion de transitorios de presion en reactores tipo BWR usando el codigo BWRDYN

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez P, J.A. [ESIME, Unidad Profesional Azcapotzalco, Av. de las Granjas 682, 02550 Mexico D.F. (Mexico)]. e-mail: jrodriguez@ipn.mx

    2007-07-01

    Several simulations of pressure transitory for a nucleo electric power station with BWR/4 type reactor were carried out. The simulated pressure transitories were made for the Peach Bottom 2 Nucleo electric central. Also, it was carried out for the same Plant the simulation of the turbine shot with derivation to the main condenser, of the reference case (benchmark) outlined by the Organization for the Cooperation and the Economic Development and of the Commission Regulatory in Nuclear matter of the United States of America. As tool to carry out the simulations of the transitory ones, the BWRDYN code developed by the Japan Energy Research Institute was used. Among the main suppositions and models that it includes the BWRDYN code its can be mentioned: a) that of punctual kinetics that calculates the neutron flow; for the calculation of the fuel temperature, this it is divided in nodes in the radial and axial directions, the wrapper is considered like a region in the radial direction; c) the pressure is supposed that it is uniform inside the reactor vessel; and d) the thermal hydraulic pattern of the reactor vessel is divided in five regions and the core is divided in several nodes to take into account the distribution of holes in the axial direction. The modeling of the control systems of the feeding water system is also included, of the pressure regulator and of the recirculation system. The systems of what is known as plant balance are also modeled. The numeric results of the simulations provide valuable information of the behavior of the nucleo electric central. The obtained results of the simulation of the reference case agree acceptably with the measurements data, when comparing them with the measurements made in the Peach Bottom 2 Central. The obtained results of each simulation are fundamental to evaluate the transitory one, as well as to delineate the sequence and the impact of diverse events that they happen during the same one transitory. In the case of the

  17. Analysis Applied Multivariate to the Studies of Stability in the Reactors BWR

    International Nuclear Information System (INIS)

    Presently work is presented the application of the analysis multivariate in the studies of stability of reactors BWR. For the confirmation of the applicability of the method of Hilbert Huang is used a group of series acquired neutronic during an outburst in the power station nuclear of Cofrentes. The peculiarity of the analyzed data is that they are not stationary and contaminated by the performance of other systems of the plant, for that that when applying the methods traditional autoregressive to these data, is values non realists of the DR In the work the DR is compared obtained by the methodology presented with the true DR and with the one obtained starting from the application of methods autoregressive to the original sign. The conclusion is evident, the value of the DR obtained by the methodology explained in this work is next to the one True DR that the resulting DR of the application of the method AR to the original sign

  18. CAE advanced reactor demonstrators for CANDU, PWR and BWR nuclear power plants

    International Nuclear Information System (INIS)

    CAE, a private Canadian company specializing in full scope flight, industrial, and nuclear plant simulators, will provide a license to IAEA for a suite of nuclear power plant demonstrators. This suite will consist of CANDU, PWR and BWR demonstrators, and will operate on a 486 or higher level PC. The suite of demonstrators will be provided to IAEA at no cost to IAEA. The IAEA has agreed to make the CAE suite of nuclear power plant demonstrators available to all member states at no charge under a sub-license agreement, and to sponsor training courses that will provide basic training on the reactor types covered, and on the operation of the demonstrator suite, to all those who obtain the demonstrator suite. The suite of demonstrators will be available to the IAEA by March 1997. (author)

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  1. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors

    International Nuclear Information System (INIS)

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

  2. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  3. Dynamic reconstruction and Lyapunov experiments from time series data in boiling water reactors. Application to B.W.R. stability analysis

    International Nuclear Information System (INIS)

    This paper shows how to obtain Lyapunov exponents from time series data on Boiling Water Reactor (BWR) stability. In order to validate the method, these characteristic exponents are compared with the ones obtained directly from the governing equations of the dynamic system. Finally, we present a method for obtaining the stability of the B.W.R. from Lyapunov exponents and describe some other applications related to limit cycles. (Author)

  4. Criticality calculations for a spent fuel storage pool for a BWR type reactor

    International Nuclear Information System (INIS)

    In this work, the methodology for the calculation of the constant of effective multiplication for the arrangement of spent fuel assemblies in the pool of a BWR type reactor is shown. Calculations were done for the pool of spent fuel specified in FSAR and for the assemblies that is thought a conservative composition of high enrichment and without Gadolinium, giving credit to the stainless steel boxes of the frames that keep the assemblies. To carry out this simulation, RECORD and MIXQUIC codes were used. With record code, macroscopic cross sections, two energy groups, for the characteristics of the thought assemblies were obtained. Cross sections, as well as the dimensions of the frames that keep the fuel assemblies were used as input data for MIXQUIC code. With this code, criticality calculations in two dimensions were done, supposing that there is not leak of neutrons along the axial of the main line. Additional calculations, supposing changes in the temperature, distance among fuel assemblies and the thickness of the stainless steel box of the frame were done. The obtained results, including the effect in tolerances due to temperature, weight and thickness, show that the arrangement in the pool, when frames are fully charged, is subcritical by less than 5% in δK. (Author)

  5. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH

    International Nuclear Information System (INIS)

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  7. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study

    International Nuclear Information System (INIS)

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  8. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  9. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

    International Nuclear Information System (INIS)

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

  10. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    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 O2; 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

  11. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

    2003-07-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  12. Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR

    International Nuclear Information System (INIS)

    The stress corrosion cracking susceptibility was studied in sensitized, solution annealed 304 steel, and in 304-L welded with a heat treatment that simulated the radiation induced segregation, by the slow strain rate test technique, in a similar environment of a boiling water reactor (BWR), 288 C, 8 MPa, low conductivity and a electrochemical corrosion potential near 200 mV. vs. standard hydrogen electrode (She). The electrochemical noise technique was used for the detection of the initiation and propagation of the cracking. The steels were characterized by metallographic studies with optical and scanning electronic microscopy and by the electrochemical potentiodynamic reactivation of single loop and double loop. In all the cases, the steels present delta ferrite. The slow strain rate tests showed that the 304 steel in the solution annealed condition is susceptible to transgranular stress corrosion cracking (TGSCC), such as in a normalized condition showed granulated. In the sensitized condition the steel showed intergranular stress corrosion cracking, followed by a transition to TGSCC. The electrochemical noise time series showed that is possible associated different time sequences to different modes of cracking and that is possible detect sequentially cracking events, it is means, one after other, supported by the fractographic studies by scanning electron microscopy. The parameter that can distinguish between the different modes of cracking is the re passivation rate, obtained by the current decay rate -n- in the current transients. This is due that the re passivation rate is a function of the microstructure and the sensitization. Other statistic parameters like the localized index, Kurtosis, Skew, produce results that are related with mixed corrosion. (Author)

  13. 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

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  14. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    Energy Technology Data Exchange (ETDEWEB)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  16. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core

    International Nuclear Information System (INIS)

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, rΘ, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  17. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  20. 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

  1. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Wagner, K.C. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  2. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor

    International Nuclear Information System (INIS)

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

  3. An assessment of BWR [boiling water reactor] Mark-II containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

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

  5. Structural integrity and management of aging in internal components of BWR reactors; Integridad estructural y manejo del envejecimiento en componentes internos de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C.R. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico, Toluca Salazar Edo. de Mexico (Mexico)]. E-mail: craj@nuclear.inin.mx

    2004-07-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  6. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

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

  7. Use of the TRAC/BF1 code in BWR reactors instability studies

    International Nuclear Information System (INIS)

    The RETRAN and TRAC codes are examples of temporary codes that are used to analyze the stability of B.W.R. Although, in many cases, this codes present good results and predict the expected behaviour, they are very sensitive to the variations of core modeling, like for example, variations in the number of cells. This can question seriously the reliability and obviously the acceptability of the analysis done with this temporary codes. In this paper we present a work using the TRAC-BF1 code to simulate the in-phase and out-of-phase oscillations, and the influence of the chose of some parameters. (author)

  8. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  10. 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)

  11. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  12. State of the art of second international exercise on benchmarks in BWR reactors

    International Nuclear Information System (INIS)

    This is a second in series of Benchmarks based on data from operating Swedish BWRs. The first one concerned measurements made in cycles 14,15 16 and 17 at Ringhals 1 Nuclear Power Plant and addressed predictive power of analytical tools used in BWR stability analysis. Part of the data was disclosed only after participants had provided their results. This work has been published in the report: NEA/NSC/DOC(96)22, November 1996. In this report it was recognised that there is a need for better qualification of the applied noise analysis methods. A follow up Benchmark was thus proposed dedicated to the analysis of time series data and including the evaluation of both global and regional stability of Forsmarks 1 and 2 Nuclear Power Plant. In this second Benchmark have participated Forsmarks Kraftgrupp AB,NEA Nuclear Science Committee, CSN Consejo de Seguridad Nuclear and Department of Chemical and Nuclear Engineering of Polytechnic University of Valencia. (Author)

  13. Characterization of welding of AISI 304l stainless steel similar to the core encircling of a BWR reactor; Caracterizacion de soldaduras de acero inoxidable AISI 304L similares a las de la envolvente del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gachuz M, M.E.; Palacios P, F.; Robles P, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    Plates of austenitic stainless steel AISI 304l of 0.0381 m thickness were welded by means of the SMAW process according to that recommended in the Section 9 of the ASME Code, so that it was reproduced the welding process used to assemble the encircling of the core of a BWR/5 reactor similar to that of the Laguna Verde Nucleo electric plant, there being generated the necessary documentation for the qualification of the one welding procedure and of the welder. They were characterized so much the one base metal, as the welding cord by means of metallographic techniques, scanning electron microscopy, X-ray diffraction, mechanical essays and fracture mechanics. From the obtained results it highlights the presence of an area affected by the heat of up to 1.5 mm of wide and a value of fracture tenacity (J{sub IC}) to ambient temperature for the base metal of 528 KJ/m{sup 2}, which is diminished by the presence of the welding and by the increment in the temperature of the one essay. Also it was carried out an fractographic analysis of the fracture zone generated by the tenacity essays, what evidence a ductile fracture. The experimental values of resistance and tenacity are important for the study of the structural integrity of the encircling one of the core. (Author)

  14. Radial distribution of UO2 and Gd2O3 in fuel cells of a BWR Reactor

    International Nuclear Information System (INIS)

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO2 enriched in U235, and of the Gadolinium (Gd2O3). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U235, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U235, some of these rods contain a concentration of Gd2O3 and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

  15. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

    Energy Technology Data Exchange (ETDEWEB)

    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  16. Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  17. Structural integrity and management of aging in internal components of BWR reactors

    International Nuclear Information System (INIS)

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  18. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR

    International Nuclear Information System (INIS)

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-θ and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-θ, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, θ and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm2s, at a height H 4 (239.07 cm) and angle 32.236o in the core shroud and 4.00 E + 09 n/cm2s at a height H 4 and angle 35.27o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)

  19. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  1. Development of a computer code for thermal hydraulics of reactors (THOR). [BWR and PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W

    1975-01-01

    The purpose of the advanced code development work is to construct a computer code for the prediction of thermohydraulic transients in water-cooled nuclear reactor systems. The fundamental formulation of fluid dynamics is to be based on the one-dimensional drift flux model for non-homogeneous, non-equilibrium flows of two-phase mixtures. Particular emphasis is placed on component modeling, automatic prediction of initial steady state conditions, inclusion of one-dimensional transient neutron kinetics, freedom in the selection of computed spatial detail, development of reliable constitutive descriptions, and modular code structure. Numerical solution schemes have been implemented to integrate simultaneously the one-dimensional transient drift flux equations. The lumped-parameter modeling analyses of thermohydraulic transients in the reactor core and in the pressurizer have been completed. The code development for the prediction of the initial steady state has been completed with preliminary representation of individual reactor system components. A program has been developed to predict critical flow expanding from a dead-ended pipe; the computed results have been compared and found in good agreement with idealized flow solutions. Transport properties for liquid water and water vapor have been coded and verified.

  2. U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 7. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been recently expanded for BWR out-of-phase behavior. Out-of-phase oscillation is a phenomenon that occurs at BWRs. During this kind of event, half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. The HRS will be used for development and validation of stability monitoring and control techniques as part of an ongoing U.S. Department of Energy Nuclear Engineering Education and Research grant. The Penn State TRIGA reactor is used to simulate BWR fundamental mode power dynamics. The first harmonic mode power, together with detailed thermal hydraulics of boiling channels of both fundamental mode and first harmonic mode, is simulated digitally in real time with a computer. Simulations of boiling channels provide reactivity feedback to the TRIGA reactor, and the TRIGA reactor's power response is in turn fed into the channel simulations and the first harmonic mode power simulation. The combination of reactor power response and the simulated first harmonic power response with spatial distribution functions thus mimics the stability phenomena actually encountered in BWRs. The digital simulations of the boiling channels are performed by solving conservation equations for different regions in the channel with C-MEX S-functions. A fast three-dimensional (3-D) reactor power display of modal BWR power distribution was implemented using MATLAB graphics capability. Fundamental mode, first harmonic, together with the total power distribution over the reactor cross section, are displayed. Because of the large amount of computation for BWR boiling channel simulation and real-time data processing and graph generation, one computer is not sufficient to handle these jobs in the hybrid reactor simulation environment. A new three-computer setup has been

  3. Recriticality in a BWR [boiling water reactor] following a core damage event

    International Nuclear Information System (INIS)

    This report describes the results of a study conducted by Pacific Northwest Laboratory to assist the US Nuclear Regulatory Commission in evaluating the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Based on a conservative bounding analysis, this report concludes that there is a potential for recriticality in BWRs if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event will most likely not generate a pressure pulse significant enough to fail the vessel. Instead, a quasi-steady power level would result and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event. Two strategies are identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached. The first strategy involves initiating boration injection at or before the time of core reflood if the potential for control blade melting exists. The second strategy involves initiating residual heat removal suppression pool cooling to remove the heat load generated by the recriticality event and thus extend the time available for boration. 31 figs., 17 tabs

  4. Application of water jet peening on BWR reactor internals with cracked surfaces

    International Nuclear Information System (INIS)

    Water Jet Peening (WJP) has been widely applied to Japanese nuclear power plants since 1999, as an approved mechanical mitigation technique against stress corrosion cracking (SCC) on materials used for nuclear reactors. Existing laboratory experience and analysis indicated that WJP had no negative impact on pre-existing cracks, such as crack propagation during its application, although there was no field experience of applying WJP on cracked surfaces. Recently, The Japan Atomic Power Company (JAPC) and Hitachi-GE Nuclear Energy, Ltd. (HGNE) have successfully applied WJP on cracked surfaces, where the cracks left as is with the requirement that they should be inspected for sizing by UT periodically to monitor the each crack propagation according to the regulator's direction. (author)

  5. Three-dimensional simulation of hydrogen detonations in the Olkiluoto BWR reactor building

    International Nuclear Information System (INIS)

    This report describes the numerical simulations of hydrogen detonations in Olkiluoto reactor building room B.60.80 using the DET3D code. The code is developed at Forschungszentrum Karlsruhe (FZK) and uses the finite difference method based on three-dimensional Euler equations for a multicomponent reacting gas. DET3D is mainly developed for modelling of gaseous detonations initiated by a direct ignition. DDT phenomena are not treated. The initial conditions of the detonation simulation were based on previous hydrogen spreading analyses carried out with the FLUENT code. DET3D calculations continued the previous, rough estimates of shock pressure loads performed with a simple DETO code. The DETO analyses were based on the strong ignition theory with oblique and normal reflection relations on the adiabatic shock waves. Shock waves were induced by point-like energy release without modelling of the propagating combustion front. In the DETO modelling, only the first shock reflection was treated. The approach of the DET3D code enables the more detailed assessment of detonation pressure loads in a real 3-D geometry. The objective of the work was to assess the pressure loads on room structures under detonation conditions. The initial conditions of detonation simulation were based on the previous hydrogen spreading analyses performed with the FLUENT code. Two sizes of leakage from the containment to the reactor building were considered: 2 mm2, which corresponds to the nominal leakage of containment, and a large leak of 20 mm2. The DET3D simulation indicated that the highest pressure spikes occurred in the room corners due to reflections and superposition of the shock waves. The highest pressure maximum in all simulation cases was about 10.6 MPa. This value was obtained in the upper corner of the room beside the containment wall. The highest pressure impulses to structures during the 150 ms simulation were about 30 - 35 kPa-s. (au)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  7. BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  8. Decay profiles of β and γ for a radionuclide inventory in equilibrium cycle of a BWR type reactor

    International Nuclear Information System (INIS)

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

  9. Calculation of the neutron flux and fluence in the covering of the nucleus and the vessel of a BWR; Calculo del flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor nuclear BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: evalle@esfm.ipn.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    One of the main objectives related with the safety in any nuclear power plant, including the nuclear power plant of Laguna Verde, is to guarantee the structural integrity of the pressure vessel of the reactor. To identify and quantifying the damage caused be neutron irradiation in the vessel of any nuclear reactor, is necessary to know as much the neutron flux as the fluence that it has been receiving during their time of operation life, since the observables damages by means of tests mechanics are products of micro-structural effects, induced by neutron irradiation, therefore, is important the study and prediction of the neutron flux to have a better knowledge of the damage that are receiving these materials. In our calculation the code DORT was used, which solves the transport equation in discreet coordinates and in two dimensions (x-y, r-{theta} and r-z), in accord to the regulator guide, it requires to make and approach of the neutron flux in three dimensions by means of the Synthesis Method. Whit this method is possible to achieve a representation of the flux in 3D combining or synthesizing the calculated fluxes by DORT code in r-{theta}, r-z and r. In this work the application of the Synthesis Method is presented, according to the Regulator Guide 1.190, to determine the fluxes 3D in the interns of a BWR using three different space meshes. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  11. 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).

  12. Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR; Determinacion de esfuerzos originados por fluctuacion de carga acustica en los secadores de vapor de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A., E-mail: javcuami26@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

  13. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors

    International Nuclear Information System (INIS)

    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)

  14. Study of instabilities in phase by using the tool {sup D}ynamics{sup :} analysis of the evolution space temporary of the waves of density in channels of reactors BWR; Estudio de las Inestabilidades en Fase Mediante la Herramienta Dinamics: analisis de la Evolucion Espacio Temporal de las Ondas de Densidad en Canales de Reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Escriva, R.; Merino, R.; Melara, J.

    2013-07-01

    This paper presents the basics of Dynamics V2 to code It allows calculations of stability for oscillations in phase in BWR reactors in the time domain. The equations of the model are exposed and is the integration of the equations. The model can be used in a large number of nodes thrust for the calculations to an acceptable computational cost, it has simplified dynamics of recirculation loop and the code has been incorporated the Oscillation in phase boundary conditions. The code incorporates the equations of boiling sub-cooled which allows to make more realistic calculations as well as subroutines to calculate the subroutines-based properties of the MATPRO and ASME.

  15. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  16. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm{sup 2}); Comportamiento a la fractura de un acero inoxidable AISI 304 sensibilizado en condiciones de reactor BWR (288 grados Centigrados y 80 Kg/cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm{sup 2}). (Author)

  17. Safety evaluation of liquid radioactive effluents treatment system in a BWR reactor, through the LIQM03 code

    International Nuclear Information System (INIS)

    In this work we made a safety evaluation of the liquid radioactive effluents system in a plant using a BWR similar to that now installed in Laguna Verde. For that purpose, the computation program ORIGENwas modified, in order to keep up to date and adapt it to the PDP 10 computer, which is operating at the Computation Department of the Nuclear Center of Mexico, the code LIQM03 was the result of this modification. As usual in this work we dealt with problems which were solved opportunely, now we have at our disposal the code LIQM03 which will be in the future a very useful tool for this kind of evaluations. (author)

  18. Technical report on material selection and processing guidelines for BWR [boiling water reactor] coolant pressure boundary piping: Final report

    International Nuclear Information System (INIS)

    This report provides the technical bases for the NRC staff's revised recommended methods to control the intergranular stress corrosion cracking susceptibility of BWR piping. For piping that does not fully comply with the material selection, testing, and processing guideline combinations of this document, varying degrees of augmented inservice inspection are recommended. This revision also includes guidance and NRC staff recommendations (not requirements) regarding crack evaluation and weld overlay repair methods for long-term operation or for continuing interim operation of plants until a more permanent solution is implemented

  19. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors; Actualizacion de los costos de combustible nuclear de la recarga de equilibrio de los reactores ABWR y EPR

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, R.F. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: rortega@fi-b.unam.mx

    2008-07-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U{sub 3O}8 in January, 2005 to a maximum of US$137.00 dollars by Ib U{sub 3}O{sub 8} by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U{sub 3}O{sub 8} such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  20. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  1. Development and Validation of NODAL-LAMBDA Program for the Calculation of the Sub-criticality of LAMDA MODES By Nodal Methods in BWR reactors

    International Nuclear Information System (INIS)

    We have developed a 3D code with two energy groups and diffusion theory that is capable of calculating eigenvalues lambda of a BWR reactor using nodal methods and boundary conditions that calculates ALBEDO NODAL-LAMBDA from the properties of the reflector code itself. The code calculates the sub-criticality of the first harmonic, which is involved in the stability against oscillations reactor out of phase, and which is needed for calculating the decay rate for data out of phase oscillations. The code is very fast and in a few seconds is able to make a calculation of the first eigenvalues and eigenvectors, discretized solving the problem with different matrix elements zero. The code uses the LAPACK and ARPACK libraries. It was necessary to modify the LAPACK library to perform various operations with five non-diagonal matrices simultaneously in order to reduce the number of calls to bookstores and simplify the procedure for calculating the matrices in compressed format CSR. The code is validated by comparing it with the results for SIMULATE different cases and making 3D BENCHMAR of the IAEA. (Author)

  2. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  3. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR

    International Nuclear Information System (INIS)

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  4. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

    2012-07-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

  5. Study and characterization of noble metal deposits on similar rusty surfaces to those of the reactor U-1 type BWR of nuclear power station of Laguna Verde

    International Nuclear Information System (INIS)

    In the present investigation work, were determined the parameters to simulate the conditions of internal oxidation reactor circulation pipes of the nuclear power plant of Laguna Verde in Veracruz. We used 304l stainless steel cylinders with two faces prepared with abrasive paper of No. 600, with the finality to obtain similar surface to the internal circulation piping nuclear reactor. Oxides was formed within an autoclave (Autoclave MEX-02 unit B), which is a device that simulates the working conditions of the nuclear reactor, but without radiation generated by the fission reaction within the reactor. The oxidation conditions were a temperature of 280 C and pressure of 8 MPa, similar conditions to the reactor operating in nuclear power plant of Laguna Verde in Veracruz, Mexico (BWR conditions), with an average conductivity of 4.58 ms / cm and 2352 ppb oxygen to simulate normal water chemistry NWC. Were obtained deposits of noble metal oxides formed on 304l stainless steel samples, in a 250 ml autoclave at a temperature range of 180 to 200 C. The elements that were used to deposit platinum-rhodium (Pt-Rh) with aqueous Na2Pt (OH)6 and Na3Rh (NO2)6, Silver (Ag) with an aqueous solution of AgNO3, zirconium (Zr) with aqueous Zr O (NO3) and ZrO2, and zinc (Zn) in aqueous solution of Zn (NO3)2 under conditions of normal water chemistry. Also there was the oxidation of 304l stainless steel specimens in normal water chemistry with a solution of Zinc (Zn) (NWC + Zn). Oxidation of the specimens in water chemistry with a solution of zinc (Zn + NWC) was prepared in two ways: within the MEX-02 autoclave unit A in a solution of zinc and a flask at constant temperature in zinc solution. The oxides formed and deposits were characterized by scanning electron microscopy, energy dispersive X-ray analysis, elemental field analysis and X-ray diffraction. By other hand was evaluated the electrochemical behavior of the oxides formed on the surface of 304l stainless steel in normal water

  6. Analysis by the Monte Carlo method of doses around the pool of storage of the control rods irradiated in a BWR reactor; Analisis mediante el metodo de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactror BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J.; Gallardo, S.

    2011-07-01

    The control rods of a boiling water reactor (BWR) are subject to a neutron flux and thus become activated during their stay in the reactor core. Activation occurs especially in the stainless steel components and impurities. The activity generated results in a dose around the bar, while it le unimportant in the reactor, but to be taken into account when removed f ron it. The bars drawn are stored on hangers placed in the storage pools of spent fuel f ron the plant. Each hanger 12 accommodates control rods and are arranged so that at least three meters of water abode the heads of the control rods. The dose received by potentially exposed workers who are in the vicinity of the storage must be calculated to ensure adequate protection of the came. This dose can be decreased significantly by changing the arrangement of the bars on hangers.

  7. Advances in the development and validation of CFD-BWR, a two-phase computational fluid dynamics model for the simulation of flow and heat transfer in boiling water reactors

    International Nuclear Information System (INIS)

    This paper presents recent advances in the validation of an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel bundles. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. We have described the model development strategy that has been adopted by the development team for the prediction of boiling flow regimes in a BWR fuel bundle. This strategy includes the use of local flow topology maps and flow topology specific phenomenological models. The paper reviews the key boiling phenomenological models and focuses on recent results of experiment analyses for the validation of two-phase BWR phenomena models including cladding-to-coolant heat transfer and Critical Heat Flux experiments and the BWR Full-size Assembly Boiling Test (BFBT). The two-phase flow models implemented in the CFD-BWR code can be grouped into three broad categories: models describing the vapor generation at the heated cladding surface, models describing the interactions between the vapor and the liquid coolant, and models describing the heat transfer between the fuel pin and the two-phase coolant. These models have been described and will be briefly reviewed. The boiling model used in the second generation of the CFD-BWR code includes a local flow topology map which allows the cell-by-cell selection of the local flow topology. Local flow topologies can range from a bubbly flow topology where the continuous phase is liquid, to a transition flow topology, to a droplet flow topology where the continuous phase is vapor, depending primarily on the local void fraction. The models describing the cladding-to-coolant heat transfer and the interplay between these models and the local flow topology are important in Critical Heat Flux (CHF) analyses, and will

  8. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel; Diseno de un nucleo de equilibrio de un reactor tipo BWR basado en un combustible de Torio-Uranio

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J.L.; Nunez C, A. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria-UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)

    2003-07-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  9. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

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

  10. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core; Determinacion de la fluencia neutronica en las soldaduras del 'core shroud' del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C. [ININ, 52750 Ocoyoacac, Estado de mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2006-07-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, r{theta}, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  11. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor

    International Nuclear Information System (INIS)

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

  12. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.

  13. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  14. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  15. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  16. Study and characterization of noble metal deposits on similar rusty surfaces to those of the reactor U-1 type BWR of nuclear power station of Laguna Verde; Estudio y caracterizacion de depositos de metales nobles sobre superficies oxidadas similares a las del reactor de la Central de Laguna Verde (CNLV) U1 del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Flores S, V. H.

    2011-07-01

    In the present investigation work, were determined the parameters to simulate the conditions of internal oxidation reactor circulation pipes of the nuclear power plant of Laguna Verde in Veracruz. We used 304l stainless steel cylinders with two faces prepared with abrasive paper of No. 600, with the finality to obtain similar surface to the internal circulation piping nuclear reactor. Oxides was formed within an autoclave (Autoclave MEX-02 unit B), which is a device that simulates the working conditions of the nuclear reactor, but without radiation generated by the fission reaction within the reactor. The oxidation conditions were a temperature of 280 C and pressure of 8 MPa, similar conditions to the reactor operating in nuclear power plant of Laguna Verde in Veracruz, Mexico (BWR conditions), with an average conductivity of 4.58 ms / cm and 2352 ppb oxygen to simulate normal water chemistry NWC. Were obtained deposits of noble metal oxides formed on 304l stainless steel samples, in a 250 ml autoclave at a temperature range of 180 to 200 C. The elements that were used to deposit platinum-rhodium (Pt-Rh) with aqueous Na{sub 2}Pt (OH){sub 6} and Na{sub 3}Rh (NO{sub 2}){sub 6}, Silver (Ag) with an aqueous solution of AgNO{sub 3}, zirconium (Zr) with aqueous Zr O (NO{sub 3}) and ZrO{sub 2}, and zinc (Zn) in aqueous solution of Zn (NO{sub 3}){sub 2} under conditions of normal water chemistry. Also there was the oxidation of 304l stainless steel specimens in normal water chemistry with a solution of Zinc (Zn) (NWC + Zn). Oxidation of the specimens in water chemistry with a solution of zinc (Zn + NWC) was prepared in two ways: within the MEX-02 autoclave unit A in a solution of zinc and a flask at constant temperature in zinc solution. The oxides formed and deposits were characterized by scanning electron microscopy, energy dispersive X-ray analysis, elemental field analysis and X-ray diffraction. By other hand was evaluated the electrochemical behavior of the oxides

  17. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code; Evaluacion del desempeno termomecanico de barras de combustible de un reactor BWR durante una rampa de potencia utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R.

    2010-07-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  18. 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

  19. Radial distribution of UO{sub 2} and Gd{sub 2}O{sub 3} in fuel cells of a BWR Reactor; Distribucion radial de UO{sub 2} y Gd{sub 2}O{sub 3} en celdas de combustible de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia del C, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62500 (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2008-07-01

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO{sub 2} enriched in U{sup 2}35, and of the Gadolinium (Gd{sub 2}O{sub 3}). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U{sup 2}35, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U{sup 2}35, some of these rods contain a concentration of Gd{sub 2}O{sub 3} and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  1. 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

  2. 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

  3. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  4. 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

  5. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  6. 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

  7. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  8. 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

  9. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  10. 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.

  11. Neutron noise analysis techniques in nuclear power reactors

    International Nuclear Information System (INIS)

    The main techniques used in neutron noise analysis of BWR and PWR nuclear reactors are reviewed. Several applications such as control of vibrations in both reactor types, determination of two phase flow parameters in BWR and stability control in BWR are discussed with some detail. The paper contains many experimental results obtained by the main author of this paper. (author)

  12. Estimating boiling water reactor decommissioning costs. A user`s manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest Lab., Richland, WA (United States)

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  13. Estimating boiling water reactor decommissioning costs. A user's manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    International Nuclear Information System (INIS)

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user's manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning

  14. 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)

  15. 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

  16. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  17. CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR usando los codigos OpenFOAM y GasFlow

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.

    2014-07-01

    using a limited number of semi-empirical data, and instead, mathematical relationships are used taking into account the various physical phenomena as well the interactions that occur among them, such as heat transfer between the fluid and the solid walls condensation of water vapor on the walls, the turbulent effects in areas of restricted passage, etc. Taking into account these advantages, this study presents a qualitative and quantitative comparison between the CFD codes OpenFOAM and Gas-Flow related to the transport phenomena of Hydrogen and other gases in the primary containment of a BWR reactor. Gas-Flow is a code of commercial license that is well validated, developed in Germany to analyze the transport of gases in nuclear reactor containments. On the other hand, OpenFOAM is an open source CFD code offering several solvers for different phenomena assessments, in this work, the reacting Foam solver is used because it has a strong similarity to the intended application of Hydrogen transport. In this thesis the results obtained using the reacting Foam solver of OpenFOAM for the calculation of transport of Hydrogen are compared with the results of the Gas-Flow code in order to assess if it is feasible to use the open source code OpenFOAM in the case of Hydrogen transport in primary containment of a BWR reactor. Some differences in the qualitative and quantitative results from both codes were found, the differences (with a maximum error rate of 4%) in the quantitative results were found are small and are considered more than acceptable for this type of analysis, moreover, these differences are mainly attributed to the transport models used, mainly because OpenFOAM uses a homogeneous mixture model and Gas-Flow a heterogeneous one. Implementing appropriate solvers in codes like OpenFOAM has the goal to develop own tools that are applicable to the transport of Hydrogen in the primary containment of a BWR reactor and thus, to gain some independence while not relying on

  18. 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)

  19. 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)

  20. 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)

  1. 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)

  2. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  3. 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)

  4. Tendencias organizacionales de las empresas.

    Directory of Open Access Journals (Sweden)

    Derlisiret Rincón M.

    2002-01-01

    Full Text Available El propósito de este trabajo es analizar la evolución de los modelos organizacionales utilizados por las empresas durante los siglos XX y XXI, a la vez que se intenta establecer las posibles tendencias para las empresas del futuro. Para ello, se caracterizan los modelos con base en la clasificación correspondiente a las visiones: mecanicista, naturista y ecologista de la organización. En tal sentido, se realiza una investigación descriptiva-documental a partir de la revisión de importantes aportes sobre las nuevas formas organizacionales, evolución y principios de las empresas. Los resultados revelan que la supervivencia de una empresa en contextos turbulentos como los actuales dependerá de variables como: rapidez, calidad, eficiencia, espontaneidad, desestandarización, flexibilidad e innovación. Se concluye que la tendencia es a la ausencia de estructuras organizacionales, formales y rígidas.

  5. Measurements and calculation of the activation of the biologic shield of the Lingen BWR power reactor definitively stopped (in view of dismantling)

    International Nuclear Information System (INIS)

    For the dismantling planning of a power reactor, it is important to know among others the depth of activation of the biological shield. A large sampling and measurement program joint to computer calculations, has given data which will allow to avoid in the future high-cost measurement programs. One shows that the calculation of activation induced by neutrons in the median plane of the core, to determine the zone from which concrete is only slightly activated. In the reactor considered, this zone does not reach the external concrete (or first layer of concrete)

  6. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume III. Checkout applications. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Obenchain, C. F.; Ramsthaler, J. H.; Eales, E. P.; Charlton, T. R.; Childs, F. W.; Giles, M. M.; Good, E. G.; Gruen, G. E.; Guttman, J.; Johnsen, G. W.; Katsma, K. R.; Keeler, C. D.; Lawford, T. W.; Mohr, C. M.; Singer, G. L.; Townsend, W. C.

    1976-09-01

    Checkout problems presented include the following: PWR large cold leg break; PWR small cold leg break; PWR intermediate sized cold leg break; BWR large recirculation line break; BWR small recirculation line break; INEL Semiscale small cold leg break; INEL LOFT large cold leg break and INEL Semiscale large cold leg break. Also included is Update 2 of the RELAP 4/M0D5 code.

  7. 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

  8. 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

  9. 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

  10. 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.

  11. Development and application of a semi-quantitative RCM approach to the reactor and Turbine Closed Cooling Water Systems in BWR

    International Nuclear Information System (INIS)

    Research and development is being performed at Chubu Electric Power Co., Inc. to develop a method for optimization of maintenance. In this study, a Reliability Centered Maintenance (RCM) method is developed that is more quantitative than the traditional function based RCM. This method is being applied to the Reactor (RCCW) and Turbine Closed Cooling Water (TCCW) Systems. The results of this research demonstrate the effectiveness of this approach in judging component 'criticality'. (author)

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. Statistical analysis in the design of nuclear fuel cells and training of a neural network to predict safety parameters for reactors BWR

    International Nuclear Information System (INIS)

    In this work the obtained results for a statistical analysis are shown, with the purpose of studying the performance of the fuel lattice, taking into account the frequency of the pins that were used. For this objective, different statistical distributions were used; one approximately to normal, another type X2 but in an inverse form and a random distribution. Also, the prediction of some parameters of the nuclear reactor in a fuel reload was made through a neuronal network, which was trained. The statistical analysis was made using the parameters of the fuel lattice, which was generated through three heuristic techniques: Ant Colony Optimization System, Neuronal Networks and a hybrid among Scatter Search and Path Re linking. The behavior of the local power peak factor was revised in the fuel lattice with the use of different frequencies of enrichment uranium pines, using the three techniques mentioned before, in the same way the infinite multiplication factor of neutrons was analyzed (k..), to determine within what range this factor in the reactor is. Taking into account all the information, which was obtained through the statistical analysis, a neuronal network was trained; that will help to predict the behavior of some parameters of the nuclear reactor, considering a fixed fuel reload with their respective control rods pattern. In the same way, the quality of the training was evaluated using different fuel lattices. The neuronal network learned to predict the next parameters: Shutdown Margin (SDM), the pin burn peaks for two different fuel batches, Thermal Limits and the Effective Neutron Multiplication Factor (keff). The results show that the fuel lattices in which the frequency, which the inverted form of the X2 distribution, was used revealed the best values of local power peak factor. Additionally it is shown that the performance of a fuel lattice could be enhanced controlling the frequency of the uranium enrichment rods and the variety of the gadolinium rods

  18. GESTIÓN HUMANA: TENDENCIAS Y PERSPECTIVAS

    OpenAIRE

    JUAN GUILLERMO SALDARRIAGA RÍOS

    2008-01-01

    En este artículo se hace referencia a las tendencias y las perspectivas de gestión humana que se imponen en el mundo en la actualidad y que, a su manera, pretenden optimizar la administración del personal de la organización y contribuir al desarrollo e incremento de la productividad y la competitividad. Mediante la realización de un Estado del Arte se logran determinar algunas de las tendencias más relevantes en la actualidad y se concluye que, cada vez con mayor fuerza, dichas tendencias se ...

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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 (

  4. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    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)

  5. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  8. GESTIÓN HUMANA: TENDENCIAS Y PERSPECTIVAS

    Directory of Open Access Journals (Sweden)

    JUAN GUILLERMO SALDARRIAGA RÍOS

    2008-01-01

    Full Text Available En este artículo se hace referencia a las tendencias y las perspectivas de gestión humana que se imponen en el mundo en la actualidad y que, a su manera, pretenden optimizar la administración del personal de la organización y contribuir al desarrollo e incremento de la productividad y la competitividad. Mediante la realización de un Estado del Arte se logran determinar algunas de las tendencias más relevantes en la actualidad y se concluye que, cada vez con mayor fuerza, dichas tendencias se sustentan en discursos que pretenden "rescatar" al ser humano dentro de la organización, lo que no necesariamente se traduce en los procesos de gestión humana que se realizan en las organizaciones nacionales e internacionales.

  9. 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)

  10. 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

  11. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  12. 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)

  13. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Jauregui Ch, V.

    2013-07-01

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

  15. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions.

  16. 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.

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  4. 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.

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  7. 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.

  8. 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

  9. 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.

  10. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

    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.

  11. BWR chromium chemistry

    International Nuclear Information System (INIS)

    This report addresses the concern about higher total specific conductivity in the reactor recirculation loop water due to the chromate ion. This concern is particularly high at plants where all other ionic species have been reduced through careful attention to makeup and condensate polisher operations. An EPRI Chromate Workshop was held in November 1990 to consider the issues raised by observed levels of chromate ion (generally 5 to 50 ppB). While BWRs on normal water chemistry were the only ones observing chromate, even plants on hydrogen water chemistry (HWC) observe sharp spikes of conductivity due to chromate whenever the hydrogen supply was interrupted after a reasonably long HWC operational period. The consensus of the workshop attendees was that chromate was not a concern as an agent causing pipe cracking compared to the more common species such as chloride and sulfate. However, the data are somewhat ambiguous for levels of chromate above 50 ppB. Adjustments to the weighing factors for the various ionic species in the industry chemistry performance index are suggested to allow for the known relative higher aggressiveness of other species relative to that of chromate

  12. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

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

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico); Castillo-Duran, Rogelio, E-mail: rogelio.castillo@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico); Palacios-Hernandez, Javier C., E-mail: javier.palacios@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico)

    2011-05-15

    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

  14. Aplicaciones Normas APA y sus tendencias

    OpenAIRE

    Ledesma Ayora, Marco; Muñoz Cajilima, Néstor Xavier

    2016-01-01

    [ES] Este manual fue elaborado en base a las normas APA (American Psychological Association), actualización y el aporte como tendencias, el anhelo es que sea de beneficio para estudiantes, docentes e investigadores que aplican una buena escritura académica. Las normas APA es un conjunto de referencias para la escritura de trabajos, proyectos y publicaciones, proyectando así su validez y confiabilidad, que permiten visibilizar una correcta formalidad en la presentación, pudiendo ser mediado a ...

  15. 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)

  16. 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)

  17. 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.

  18. 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)

  19. Tendencias francesas en materia de dolo

    Directory of Open Access Journals (Sweden)

    Aude Denizot-Libreros

    2014-07-01

    Full Text Available Numerosos fallos recientes de la Corte de Casación francesa en materia de dolo han suscitado nuevas investigaciones en la materia. El dolo siempre ha sido una noción ambigua, objeto de un sinnúmero de discusiones. El propósito que se persigue consiste en ofrecer un panorama muy general de las tendencias francesas en esta materia. Estudiar estas tendencias, doctrinales y jurisprudenciales, implica confrontarse con varios temas espinosos, especialmente los de la reticencia, el dolo incidente y el dolo por parte del tercero. También se analiza el dolo con relación a otros conceptos: el error sobre el valor, el error inexcusable, y, en materia de sanción, tanto la pérdida de oportunidad como la noción de refacción del contrato. Entre debates doctrinales y vacilaciones jurisprudenciales, en este tema el derecho francés se caracteriza más por sus aproximaciones que por su rigor, delineando por contraste los impases a evitar.

  20. 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

  1. 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

  2. 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

  3. Development of a dynamic model of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    A dynamic model of a nuclear power plant, including a boiling water reactor, high- and low-pressure turbines, moisture separator, reheater, condenser, feedwater heaters and feedwater pump, was developed. The model is one-dimensional except for the nuclear part of the reactor, which is based on the point kinetics equation, and the condenser model and feedwater pump model. It has been used to study different transients occuring during normal operating conditions and for evaluating the control systems of a BWR nuclear power plant. Particular emphasis was laid on the reactor pressure control system and the recirculation flow control system. (author)

  4. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

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

  6. 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.

  7. 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)

  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. TRAC-BD1: transient reactor analysis code for boiling-water systems

    Energy Technology Data Exchange (ETDEWEB)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented.

  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. 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

  12. 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.

  13. 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)

  14. 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

  15. 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.)

  16. 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

  17. Characterization of corrosion layers on irradiated and non-irradiated surfaces in BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J.; Balek, V.; Zmitko, M.; Brozova, A.; Burda, J. [Nuclear Research Inst., Rez (Czech Republic); Hoffmann, H.; Ruehle, W. [VGB Essen (Germany); Bezdicka, P. [Institute of Inorganic Chemistry, ASCR, Rez (Czech Republic)

    2002-07-01

    Stress corrosion cracking of low-alloyed steel 22NiMoCr37 is evaluated with the goal to determine crack growth rate in irradiated steel under conditions simulating closely conditions of BWR RPV under operation. For the experiment, in pile BWR experimental loop has been built at Nuclear Research Institute, Rez. During the experiment, specimens are loaded by cyclic and constant load. Crack growth is monitored by means of potential drop measurement and COD. Corrosion layers formed on specimens in reactor water loop exposed to BWR primary water chemistry and radiation were studied. Two sets of specimens were placed in loop channels. One set of specimens was situated in reactor conditions and the second set out of reactor, other parameters like water chemistry (e.g. concentration of hydrogen, oxygen and conductivity), temperature and flow rate were identical. By means of this an effect of radiation could be studied. The differences in chemical composition, structure and microstructure of corrosion products were characterized by SEM and X-ray powder diffractometry. The differences in microstructure of corrosion layer formed under different conditions were observed. (authors)

  18. 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)

  19. 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)

  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. IVA2 - a computer code for modelling of transient 3D-three phase three component flows using three velocity fields in cylindrical geometry with arbitrary internals including nuclear reactor PWR/BWR-core

    International Nuclear Information System (INIS)

    This report contains a formal code description (description of the input data, contents of the COMMON blocks, functions of the IVA2/001 routines). In addition the nonformal description of the current IVA2/001 constitutive package and the reactor core model are given. (orig.)

  2. 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.)

  3. 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

  4. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  5. 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.)

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

    Directory of Open Access Journals (Sweden)

    Antonella Lombardi Costa

    2008-01-01

    Full Text Available 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 presented. The RELAP5/MOD3.3 thermal-hydraulic (TH system code and the PARCS-2.4 3D neutron kinetic (NK code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR and the natural frequency (NF from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs. The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.

  7. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  8. Light Water Reactor Sustainability Program BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates

    Energy Technology Data Exchange (ETDEWEB)

    Sebastien Teysseyre

    2014-04-01

    As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

  9. Condensate treatment in BWR circuits by filter demineralizer units using powdered ion exchange resin at medium and high temperature

    International Nuclear Information System (INIS)

    Considering the radiation build-up in some BWR reactors, we make a correlation between this phenomenon and the condensate purification system applied and the point of its utilization into the circuits. The application temperature of such a plant seems to have a very important role on the equilibria of metals contained in the reactor water and on the oxide composition. The efficiency of the condensate polishing system and the corrosion control are the most interesting objectives to achieve and to maintain, to control and regulate the physical and chemical process in the feedwater and in the reactor water. Up to date the technology owns major knowledge and a consistent know-how on using chemical products in order to increase the condensate polishing system efficiency. It is also considered a typical parallel case of a conventional power station and a secondary system of BWR units. (author)

  10. SIMULATE-3K: Enhancements and Application to Boiling Water Reactor Transients

    International Nuclear Information System (INIS)

    The SIMULATE-3K (S-3K) reactor analysis code has been applied to a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) transients since 1993. Over the years, many changes have occurred in the S-3K channel hydraulics and ex-core component modeling. This paper summarizes those changes and outlines the status of existing vessel and steam line models. Examples are given for BWR transients that can be analyzed with S-3K

  11. 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)

  12. 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

  13. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde; Rango del monitor de radiacion para el venteo rigido de la contencion primaria durante operacion normal y emergencia para un reactor BWR-5 en Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  14. 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

  15. 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

  16. 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.

  17. 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

  18. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-09-01

    This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format.

  19. Proposal for the construction of a high resolution tomography installation in the BWR-type reactor Dodewaard, Netherlands. Voorstel voor de bouw van een hoge resolutie tomografie-opstelling in de centrale Dodewaard

    Energy Technology Data Exchange (ETDEWEB)

    Jaspers, J.E.; Tanke, R.H.J. (KEMA Onderzoek en Ontwikkeling, Arnhem (Netherlands))

    1993-01-01

    KEMA carried out several measurements on fission fuel rods and fission fuel elements of the nuclear reactor Dodewaard within the framework of the project Tomography Fission Fuel. The use of tomography provides information on the integrity of fission fuel, the burnup, the diffusion of fission fuel gases and the fission fuel/cladding interaction. It appeared that the present installation is not appropriate for high resolution tomography. Therefore it has been decided to design and construct a new tomography installation. Also attention is paid to the estimation of the costs for the chosen installation. The results of the title study can be used for the inspection of fission fuel elements in nuclear power plants and for local burnup analysis in fission fuel elements.

  20. Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Williams, T. [EGL Laufenburg (Switzerland); Helmersson, S. [Westinghouse Atom (Sweden)

    2001-03-01

    The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. El Ahorro Privado en Venezuela: Tendencias y Determinantes

    OpenAIRE

    Matías Riutort; Rafael Muñoz; Juan Carlos Guevara; Luis Zambrano Sequín

    1998-01-01

    La característica notable del comportamiento de la tasa de ahorro privado en Venezuela es su tendencia secular a reducirse desde los elevados niveles experimentados durante los años setenta a los muy reducidos de los noventa. Adicionalmente a esta tendencia al deterioro, habría que añadir su alta volatilidad. Las razones que explican este resultado no sólo están referidas a los sucesivos shocks, externos e internos, que han afectado a Venezuela, sino también a la forma peculiar como se ha ges...

  6. 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

  7. 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.

  8. Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR; Foerstudie av stroemningsinducerade laster paa interndelar vid brott i huvudcirkulationskretsarna i BWR

    Energy Technology Data Exchange (ETDEWEB)

    Marcinkiewicz, Jerzy; Lindgren, Anders [Det Norske Veritas Nuclear Technology AB, Stockholm (Sweden)

    2002-12-01

    Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report.

  9. 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

  10. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Science.gov (United States)

    Tanaka, Ken-ichi

    2016-06-01

    We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV) of a Boiling Water Reactor (BWR) by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au) and Nickel (Ni) at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  11. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

    Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  12. 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.)

  13. Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR

    International Nuclear Information System (INIS)

    Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report

  14. Aircraft-crash-protected steel reactor building roof structure for the European market

    International Nuclear Information System (INIS)

    This paper recommends the use of all steel roof structures for the reactor building of European Boiling Water Reactor (BWR) plants. This change would make the advanced US BWR designs more compatible with European requirements. Replacement of the existing concrete roof slab with a sufficiently thick steel plate would eliminate the concrete spelling resulting from a postulated aircraft crash, potentially damaging the drywell head or the spent fuel pool

  15. Aircraft-crash-protected steel reactor building roof structure for the European market

    Energy Technology Data Exchange (ETDEWEB)

    Posta, B.A.; Kadar, I. [Bechtel Corp., San Francisco, CA (United States); Rao, A.S. [General Electric Nuclear Engineering, San Jose, CA (United States)

    1996-07-01

    This paper recommends the use of all steel roof structures for the reactor building of European Boiling Water Reactor (BWR) plants. This change would make the advanced US BWR designs more compatible with European requirements. Replacement of the existing concrete roof slab with a sufficiently thick steel plate would eliminate the concrete spelling resulting from a postulated aircraft crash, potentially damaging the drywell head or the spent fuel pool.

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

    International Nuclear Information System (INIS)

    It is the purpose of this report to set forth acceptable methods to reduce the stress corrosion cracking susceptibility of BWR piping and thereby also provide an increased level of reactor coolant pressure boundary integrity. Because the most straightforward and desirable approach or methods may not be practicable, or even possible, for all plants, the bases for varying degrees of conformance to the guidelines are provided. Augmented inservice inspection and leak detection requirements are established for plants that have not fully implemented the provisions presented

  17. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cheng L. Y.; Baek J.; Cuadra,A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  18. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR

    International Nuclear Information System (INIS)

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  19. Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

    Energy Technology Data Exchange (ETDEWEB)

    Su, T.M.

    1982-10-01

    Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report.

  20. 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

  1. Overview of activities for the reduction of dose rates in Swiss boiling water reactors

    International Nuclear Information System (INIS)

    Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

  2. 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)

  3. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  4. BWR stability analyses at BNL

    International Nuclear Information System (INIS)

    The March 9, 1988 instability at the LaSalle County-2 boiling water reactor power plant at Seneca, IL was simulated with Brookhaven National Laboratory's (BNL's) Engineering Plant Analyzer (EPA) for the purpose of demonstrating that the EPA is suitable for simulating large-amplitude, limit-cycle power and flow oscillations. It was shown in fall of 1988, by comparing all the available plant data from the STARTREC recording system of LaSalle-2 with EPA simulation results, that the EPA reproduces the LaSalle-2 oscillations without the use of stabilizing or destabilizing model or parameter modifications. The power vs. flow map of the LaSalle-2 plant was also reproduced at five lines of constant control rod positions. The LaSalle-2 stability boundary was established with the EPA and confirmed within ±15% accuracy by comparing the EPA results with the results of the frequency domain code LAPUR of Oak Ridge National Laboratory. Comparisons of EPA simulation results with plant data from three Peach Bottom stability tests show an agreement, based on mean and standard deviation, of -10±28%, -1±40% and +28±52% (low power) in the gain of the pressure to power transfer functions. This demonstrates that the time domain code HIPA in the EPA is capable of simulating instabilities

  5. CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy (Finland); Bjoere, S.; Olsson, Lena [ABB Atom AB, Vaesteraas (Sweden)

    1992-12-01

    Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems.

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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.

  11. 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

  12. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F

    2000-07-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  13. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    Science.gov (United States)

    Ratnayake, Ruwan Kumara

    While best estimate computer codes provide the licensing basis for nuclear power facilities, they also serve as analytical tools in overall plant and component design procedures. An ideal best estimate code would comprise of universally applicable mechanistic models for all its components. However, due to the limited understanding in these specific areas, many of the models and correlations used in these codes reflect high levels of empiricism. As a result, the use of such models is strictly limited to the range of parameters within which the experiments have been conducted. Disagreements between best estimate code predictions and experimental results are often explained by the mechanistic inadequacies of embedded models. Significant mismatches between calculated and experimental critical power values are common observations in the analyses of Boiling Water Reactors (BWR). Based on experimental observations and calculations, these mismatches are attributed to the additional entrainment and deposition caused by spacer grids in BWR fuel assemblies. In COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid); a state of the art industrial best estimate code, these disagreements are hypothesized to occur due the absence of an appropriate spacer grid model. In this thesis, development of a suitably detailed spacer grid model and integrating it to COBRA-TF is documented. The new spacer grid model is highly mechanistic so that the applicability of it is not seriously affected by geometric variations in different spacer grid designs. COBRA-TF (original version) simulations performed on single tube tests and BWR rod bundles with spacer grids showed that single tube predictions were more accurate than those of the rod bundles. This observation is understood to arise from the non-availability of a suitable spacer grid model in COBRA-TF. Air water entrainment experiments were conducted in a test section simulating two adjacent BWR sub channels to visualize the flow behavior at

  14. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Chopra, O. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Gruber, Eugene E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shack, William J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  15. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Gonzalez, Jesus S. [Univ. Politecnica de Madrid (Spain); Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William BJ J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  16. Fundamentals of nuclear power plants with light water reactors. Pt. 1

    International Nuclear Information System (INIS)

    The authors give a comprehensive picture (in two parts) of modern LWR reactors. All technical constructive and physical details of BWR and PWR reactors are described and compared. The first part describes the different cooling systems and their components, including control systems. In the second part, the layout of the reactor core, fuel assemblies, instrumentation and thermohydraulic aspects are reported on. (TK)

  17. Final results of the XR2-1 BWR metallic melt relocation experiment

    International Nuclear Information System (INIS)

    This report documents the final results of the XR2-1 boiling water reactor (BWR) metallic melt relocation experiment, conducted at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The objective of this experiment was to investigate the material relocation processes and relocation pathways in a dry BWR core following a severe nuclear reactor accident such as an unrecovered station blackout accident. The imposed test conditions (initial thermal state and the melt generation rates) simulated the conditions for the postulated accident scenario and the prototypic design of the lower core test section (in composition and in geometry) ensured that thermal masses and physical flow barriers were modeled adequately. The experiment has shown that, under dry core conditions, the metallic core materials that melt and drain from the upper core regions can drain from the core region entirely without formation of robust coherent blockages in the lower core. Temporary blockages that suspended pools of molten metal later melted, allowing the metals to continue draining downward. The test facility and instrumentation are described in detail. The test progression and results are presented and compared to MERIS code analyses. 6 refs., 55 figs., 4 tabs

  18. Electrochemical response to hydrogen water chemistry at the J.A. FitzPatrick BWR

    International Nuclear Information System (INIS)

    It was the goal of the HWC campaign at the FitzPatrick BWR to determine the hydrogen injection rates required to mitigate IGSCC and IASCC in the reactor internals. Electrochemical sensors were installed at two elevations in one of the local power range monitors (LPRMs). In the summer of 1990 the HWC campaign was conducted. The feedwater hydrogen injection rate was varied from 12 to 90 standard cubic feet/minute (SCFM) and the ECPs from the sensors in the LPRM were measured. The relationship of hydrogen injection versus ECP was determined with specific emphasis on the injection rate required to decrease the ECP to -0.230 V(SHE) at each location in the LPRM. The LPRM lower position, equivalent to the outlet of the lower plenum, required three times more hydrogen injection than previously determined for the recirculation piping system to achieve -0.230 V(SHE). The upper position in the LPRM required far greater hydrogen injection rates to approach the protection potentials. Since completion of the FitzPatrick test, a program with similar objectives was conducted at an overseas BWR. It was found that in the high radiation environment of the core bypass newly designed platinum sensors performed quite adequately as reversible reference electrodes. These results provide a possible approach for protection of key reactor structurals with minimum hydrogen injection and low main steam line dose rates

  19. Comparison of metaheuristic optimization techniques for BWR fuel reloads pattern design

    International Nuclear Information System (INIS)

    Highlights: ► This paper shows a performance comparison of several optimization techniques for fuel reload in BWR. ► Genetic Algorithms, Neural Networks, Tabu Search and several Ant Algorithms were used. ► All optimization techniques were executed under same conditions: objective function and an equilibrium cycle. ► Fuel bundles with minor actinides were loaded into the core. ► Tabu search and Ant System were the best optimization technique for the studied problem. -- Abstract: Fuel reload pattern optimization is a crucial fuel management activity in nuclear power reactors. Along the years, a lot of work has been done in this area. In particular, several metaheuristic optimization techniques have been applied with good results for boiling water reactors (BWRs). In this paper, a comparison of different metaheuristics: genetic algorithms, tabu search, recurrent neural networks and several ant colony optimization techniques, were applied, in order to evaluate their performance. The optimization of an equilibrium core of a BWR, loaded with mixed oxide fuel composed of plutonium and minor actinides, was selected to be optimized. Results show that the best average values are obtained with the recurrent neural networks technique, meanwhile the best fuel reload was obtained with tabu search. However, according to the number of objective functions evaluated, the two fastest optimization techniques are tabu search and Ant System.

  20. Final results of the XR2-1 BWR metallic melt relocation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, R.O.; Humphries, L.L.

    1997-08-01

    This report documents the final results of the XR2-1 boiling water reactor (BWR) metallic melt relocation experiment, conducted at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The objective of this experiment was to investigate the material relocation processes and relocation pathways in a dry BWR core following a severe nuclear reactor accident such as an unrecovered station blackout accident. The imposed test conditions (initial thermal state and the melt generation rates) simulated the conditions for the postulated accident scenario and the prototypic design of the lower core test section (in composition and in geometry) ensured that thermal masses and physical flow barriers were modeled adequately. The experiment has shown that, under dry core conditions, the metallic core materials that melt and drain from the upper core regions can drain from the core region entirely without formation of robust coherent blockages in the lower core. Temporary blockages that suspended pools of molten metal later melted, allowing the metals to continue draining downward. The test facility and instrumentation are described in detail. The test progression and results are presented and compared to MERIS code analyses. 6 refs., 55 figs., 4 tabs.

  1. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  2. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    Energy Technology Data Exchange (ETDEWEB)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  3. 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''

  4. 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

  5. Effect of two impurities and zinc on stress corrosion cracking of stainless steel and nickel alloys in BWR environments

    International Nuclear Information System (INIS)

    Boiling water reactors (BWRs) operate with very high purity water with only small additions of dissolved hydrogen and, most recently, noble metals. However, even operation with very low conductivity water (e.g., 0.07 μS/cm) coolant will not prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel and nickel alloys under atypical oxygenated conditions. The presence of certain impurities dissolved in the coolant can dramatically increase the propensity of this most insidious form of environmentally-assisted cracking. The goal of this paper is to present the effect of effect of chloride and sulfate plus zinc on the IGSCC propensities of BWR piping and reactor internals under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions. While it is well documented the sulfate and chloride are particularly aggressive in promoting IGSCC of BWR structural materials, several anions such as chromate and nitrate have little impact while of zinc added as zinc oxide appears to be beneficial. To emphasize the effect of impurities on the structural integrity of BWR components in perspective, the BWR fleet's most severe documented water chemistry transient, where the conductivity reached on 232 μS/cm with 21.2 ppm chloride and 93.8 ppm sulfate, will be presented. For example, on-line real-time crack growth rate measurements using the highly accurate reversing DC potential drop technique revealed a crack growth rate increase by almost a factor of 300 for an Alloy 182 weld metal compact tension fracture mechanics specimen during this raw water transient. The recommendations for subsequent plant inspection and start up after this transient will also be discussed where the value of real time crack growth rate monitoring cannot be overemphasized. (author)

  6. 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)

  7. BWR core thermal-hydraulic stability in anticipated transients without SCRAM

    International Nuclear Information System (INIS)

    The potential exists for core thermal hydraulic density wave oscillations to occur in the boiling water reactor (BWR) during anticipated transients without scram (ATWS) events. Conditions which may lead to oscillations in ATWS result either from recirculation pump trip (RPT) from appropriate ATWS signals, or from the failure to scram during core oscillations. A program to evaluate these scenarios will assess the adequacy of the system design and emergency operating procedures relative to the reactor system performance. Evaluation of ATWS events with oscillations is extremely complex, and defining the bases for the analyses is critical to the success of the program. Accurate analysis requires detailed modeling of the sequence of events, the plant configuration, and initial conditions

  8. Analysis of non-linear BWR stability behavior applying proper orthogonal decomposition

    International Nuclear Information System (INIS)

    The main drivers of BWR stability behavior are the multiple thermal hydraulic interactions between power, flow rate, and density, reinforced by the Neutronics feedback. This coupling is schematically presented in Figure 1. Especially for high power low flow operating conditions associated with unfavorable power distribution BWR operation requires attention with respect to power oscillations. Admissible reactor operation conditions maintain a certain distance to the stability limit given by linear theory. Evaluation of non-linear states requires application of time domain codes or measurement data but this depends on the specific transients considered. Improvements of non-linear stability analysis focus on the accelerating of simulations and to provide assessment for the whole parameter space. In our transient analysis, the physical behavior of the system is approximated by a reduced order model (ROM) that respects stability relevant characteristics. More precisely, the system of coupled non-linear partial differential equations (PDEs) is mapped to coupled non-linear ordinary differential equations (ODEs) that can be solved faster and analyzed with respect to non-linear stability phenomena. Proper orthogonal decomposition (POD), i.e. a spectral method based on experimental or computational fluid dynamic (CFD) data, is capable to detect oscillating states of the physical system needed. Moreover, POD provides a well-defined truncation criterion for the minimum number of modes. A standard Galerkin method employing POD modes as Ansatz functions yields a non-linear ROM. The exceptional advantage of our methodology is its generality. It is accessible for various physical systems including the reactor dynamics of BWR. We envision a fully coupled non-linear investigation of the BWR system. The method benefits from a well defined sequence of processing steps which are automated to a large extent. This minimizes the required user interaction. Obviously the user still needs to

  9. Experimental and analytical study on thermal hydraulics in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Araya, Fumimasa; Ohnuki, Akira; Yoshida, Hiroyuki; Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Study and development of reduced-moderation spectrum water reactor proceeds as a option of the future type reactor in Japan Atomic Energy Research Institute (JAERI). The reduced-moderation spectrum in which a neutron has higher energy than the conventional water reactors is achieved by decreasing moderator-to-fuel ratio in the lattice core of the reactor. Conversion ratio in the reduced-moderation water reactor can be more than 1.0. High burnup and long term cycle operation of the reactor are expected. A type of heavy water cooled PWR and three types of BWR are discussed as follows; For the PWR, (1) critical heat flux experiments in hexagonal tight lattice core, (2) evaluation of cooling limit at a nominal power operation, and (3) analysis of rewetting cooling behavior at loss of coolant accident following with large scale pipe rupture. For the BWR, analyses of cooling limit at a nominal power operation of, (1) no blanket BWR, (2) long term cycle operation BWR, and (3) high conversion ratio BWR. The experiments and the analyses proved that the basic thermal hydraulic characteristics of these reduced-moderation water reactors satisfy the essential points of the safety requirements. (Suetake, M.)

  10. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  11. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    International Nuclear Information System (INIS)

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code's capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

  14. Study of transient rod extraction failure without RBM in a BWR; Estudio del transitorio error de extraccion de barra sin RBM en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

  15. Study of transient turbine shot without bypass in a BWR; Estudio del transitorio disparo de turbina sin bypass en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    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)

  16. Crack growth rate in core shroud horizontal welds using two models for a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis Juárez, C.R., E-mail: carlos.arganis@inin.gob.mx; Hernández Callejas, R.; Medina Almazán, A.L.

    2015-05-15

    Highlights: • Two models were used to predict SCC growth rate in a core shroud of a BWR. • A weld residual stress distribution with 30% stress relaxation by neutron was used. • Agreement is shown between the measurements of SCC growth rate and the predictions. • Slip–oxidation model is better at low fluences and empirical model at high fluences. - Abstract: An empirical crack growth rate correlation model and a predictive model based on the slip–oxidation mechanism for Stress Corrosion Cracking (SCC) were used to calculate the crack growth rate in a BWR core shroud. In this study, the crack growth rate was calculated by accounting for the environmental factors related to aqueous environment, neutron irradiation to high fluence and the complex residual stress conditions resulting from welding. In estimating the SCC behavior the crack growth measurements data from a Boiling Water Reactor (BWR) plant are referred to, and the stress intensity factor vs crack depth throughout thickness is calculated using a generic weld residual stress distribution for a core shroud, with a 30% stress relaxation induced by neutron irradiation. Quantitative agreement is shown between the measurements of SCC growth rate and the predictions of the slip–oxidation mechanism model for relatively low fluences (5 × 10{sup 24} n/m{sup 2}), and the empirical model predicted better the SCC growth rate than the slip–oxidation model for high fluences (>1 × 10{sup 25} n/m{sup 2}). The relevance of the models predictions for SCC growth rate behavior depends on knowing the model parameters.

  17. Tendencias gerenciales y la gestión universitaria

    Directory of Open Access Journals (Sweden)

    Thaís Ferrer

    2004-01-01

    Full Text Available El propósito de la investigación consiste en determinar las tendencias gerenciales contemporáneas que aplica la alta gerencia universitaria de la Universidad del Zulia (LUZ, en su gestión para la toma de decisiones en las instituciones de educación superior, cuyos retos científicos y tecnológicos son cada día mayores en el contexto competitivo de las organizaciones. La investigación sigue la modalidad de carácter descriptivo aplicada. La población objeto de estudio fue clasificada como población censal, por sus características definitorias incluye a las autoridades universitarias, rectorales y decanales. Los resultados fueron analizados mediante distribución de frecuencias. Con los datos obtenidos se concluyó que la planificación estratégica se aplica en un 100%, la calidad total con un 68.4%, la reingeniería con el 31.6% y el benchmarking con el 26.3%. Las tendencias no utilizadas fueron el empowerment, la organización virtual y el cuadro de mando integral. La generación de lineamientos para la gerencia deseable de la alta gerencia universitaria, incluye: liderazgo, manejo de herramientas tecnológicas, capacitación profesional y gerenciar con técnicas contemporáneas. Se determina que sí existe viabilidad institucional para aplicar las propuestas. La investigación arrojó aportes significativos rompiendo el paradigma de investigar la alta gerencia de LUZ, generando el perfil de la gerencia deseable, recomendándose la aplicación de las tendencias gerenciales contemporáneas ajustándose a la naturaleza y momento presente y futuro de la organización.

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. Semi-automated proper orthogonal decomposition reduced order model non-linear analysis for future BWR stability

    International Nuclear Information System (INIS)

    Highlights: • Techniques within the field of ROMing based on POD are reviewed regarding “well-behaved” applications. • A systematic, general, mostly automated, reduction methodology based on POD is derived. • It is applicable for many classes of dynamical problems including the envisioned BWR application. • Robustness of this approach is demonstrated by a “pathological” test example. • The derived ROM accurately predicts dynamics of transients not included in the data set. - Abstract: Thermal–hydraulic coupling between power, flow rate and density, intensified by neutronics feedback are the main drivers determining the stability behavior of a boiling water reactor (BWR). High-power low-flow conditions in connection with unfavorable power distributions can lead the BWR system into unstable regions where power oscillations can be triggered. This important threat to operational safety requires careful analysis for proper understanding. Current design rules assure admissible operation conditions by exclusion regions determined by numerical calculations and analytical methods based on non-linear states for specific transients. Analyzing an exhaustive parameter space of the non-linear BWR system becomes feasible with methodologies based on reduced order models (ROMs), saving computational cost and improving the physical understanding. A new self-contained methodology is developed, based on the general general proper orthogonal decomposition (POD) reduction technique. It is mostly automated, applicable for generic partial differential equation (PDE) systems, and reduces them in a grid-free manner to a small ordinary differential equation (ODE) system able to capture even non-linear dynamics. This allows a much more extensive analysis of the represented physical system. Symbolic mathematical manipulations are performed automatically by Mathematica routines. A novel and general calibration roadmap is proposed which simplifies choices on specific POD

  3. Development of Sirius facility that simulates void-reactivity feedback, and regional and core-wide stability estimation of natural circulation BWR

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, M.; Inada, F.; Yasuo, A. [Tokyo Electric Power Co., Inc., Central Research Institute (Japan)

    2001-07-01

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of fluid conditions, power distributions, and fuel rod thermal conductivity time constants, including the normal operating conditions of a typical natural circulation BWR. The results showed that there is a sufficiently wide stability margin under normal operating conditions, even when void-reactivity feedback is taken into account. (author)

  4. Verification of the Advanced Nodal Method on BWR Core Analyses by Whole-Core Heterogeneous Transport Calculations

    International Nuclear Information System (INIS)

    Recent boiling water reactor (BWR) core and fuel designs have become more sophisticated and heterogeneous to improve fuel cycle cost, thermal margin, etc. These improvements, however, tend to lead to a strong interference effect among fuel assemblies, and it my cause some inaccuracies in the BWR core analyses by advanced nodal codes. Furthermore, the introduction of mixed-oxide (MOX) fuel will lead to a much stronger interference effect between MOX and UO2 fuel assemblies. However, the CHAPLET multiassembly characteristics transport code was developed recently to solve two-dimensional cell-heterogeneous whole-core problems efficiently, and its results can be used as reference whole-core solutions to verify the accuracy of nodal core calculations. In this paper, the results of nodal core calculations were compared with their reference whole-core transport solutions to verify their accuracy (in keff, assembly power and pin power via pin power reconstruction) of the advanced nodal method on both UO2 and MOX BWR whole-core analyses. Especially, it was investigated if there were any significant differences in the accuracy between MOX and UO2 results

  5. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    International Nuclear Information System (INIS)

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  6. Operating experience of reactors points up need for new thermal-hydraulic inquiries

    International Nuclear Information System (INIS)

    Review of accident and preaccident situation in the context of thermal-hydraulic processes in PWR and BWR is presented. The most frequently occurring preaccident events in the reactor operation pertaining to thermal-hydraulic processes: water hammer, thermal fatigue, transition processes, supercooling, formation of vortex, oscillation of power in BWR are discussed. Activation of theoretical and experimental thermal-hydraulic studies with the aim of improvement of safety and efficiency of NPU is proposed

  7. Nuclear coupled flow instability study for natural circulation BWR startup transient

    International Nuclear Information System (INIS)

    Natural circulation Boiling Water Reactor (BWR) startup transient was investigated in Purdue University Multidimensional Test Assembly (PUMA) facility based on a natural circulation BWR design. Strategy and results of the experiments, which consider the effects of void-reactivity and fuel heat conduction time constant, are discussed. Total reactivity is treated to be composed of two components: external reactivity due to control rod motion and void-reactivity. A detailed analysis for heat conduction problem is performed to derive dimensionless groups. Based on area-averaged heat conduction equations for pellet and clad regions, Fourier and Biot numbers are derived to simulate wall heat flux response. Power transient, which has been used for startup transient investigation without void-reactivity feedback is used to derive the control rod reactivity. Twelve conductivity probes are used to measure local void fraction inside core at three axial locations. The local void-fraction data is used to calculate volume average void fraction, which is used to calculate the voil-reactivity. A real-time Point Kinetic Model solver is implemented to PUMA heater power control program to determine power transient during startup. The results demonstrate that the inclusion of void-reactivity feedback worsen the scenario for startup instabilities and may cause large amplitude neutron flux oscillations. (author)

  8. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  9. An intermediate break BWR LOCA test (RUN 991) at ROSA-III

    International Nuclear Information System (INIS)

    Double failures on the emergency-core-cooling systems (ECCSs) can be resulted in a case of loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) by assuming an ECCS line break and the single failure criterion on another ECCS. In the Rig-of-Safety Assessment (ROSA)-III program, two BWR LOCA simulation tests with intermediate break areas were performed to experimentally study influences of the ECCS double failures on core cooling phenomena. As there was no break unit in the ROSA-III ECCS lines, two break locations were selected above and below the ECCS line elevation. Namely, one is a main steam line (MSL) break test of RUN 992 which was previously reported. Another one is a single-ended jet pump drive line (JPDL) break test of RUN 991. And this break location effect on the system responses was briefly studied in a report of JAERI 1307. This report presents precise experiment results of RUN 991 with respect to the core cooling phenomena related to transient system mass and also presents additional findings on the influences of ECCS double failures in some intermediate break LOCA tests including above two tests. (author)

  10. Thermal-hydraulics stability of natural circulation BWR under startup. Flashing effects

    International Nuclear Information System (INIS)

    To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a boiling natural circulation reactor like the ESBWR. In such systems, thermal-hydraulic stability during low pressure start-up should be examined while considering the flashing induced by the pressure drop in the channel and the chimney due to gravity head. In this work, a BWR stability analysis code in the frequency domain, named FISTAB (Flashing-Induced STability Analysis for BWR), was developed to address the issue of flashing-induced instability. A thermal-hydraulics non-homogeneous equilibrium model (NHEM) based on a drift flux formulation along with a lumped fuel dynamics model is incorporated in the work. The vapor generation rate is derived from the mixture energy conservation equation while considering the effect of flashing. The functionality of the FISTAB code was confirmed by comparison to experimental results from SIRIUS-N facility at CRIEPI, Japan. Both stationary and perturbation results agree well with the experimental results. (author)

  11. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models; SUN-RAH: simulador universitario de nucleoelectrica BWR basado en modelos de orden reducido

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2003-07-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  12. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  13. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  14. Analysis of a German BWR core with TRACE/PARCS using different cross section sets

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, C., E-mail: Christoph.Hartmann@kit.edu [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Westinghouse Electric Germany GmbH, Mannheim (Germany); Sanchez, V.H. [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Tietsch, W. [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2011-07-01

    'Full text:' Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools as well as appropriate cross sections (XS) libraries depending on the individual thermal hydraulic state parameters. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running and user-friendly lattice codes such as Casco and Helios. At research institutions and universities, alternative tools to the commercial ones with full access to the source code as well as moderate cost are urgently needed. The Scale system is being developed and improved for lattice physics calculations of real core loading of Light Water Reactors (LWR). It represents a promising alternative to the commercial lattice codes. At Karlsruhe Institute of Technology (Kit) a computational route based on Scale/Triton/Newt for BWR core loading is under development. The generated XS-data sets have to be transformed in PMAXS-format for use in the reactor dynamic code PARCS. This task is performed by the module GenPMAXS being developed and tested at the Michigan University. To verify the computational route, a BWR fuel assembly depletion problem was calculated by PARCS and compared to the CASMO results. Since the SCALE/TRITON XS-file does actually not contain all required neutronic data, FORTRAN routines have been developed to incorporate the missing data e.g. the yields of Iodine, Xenon and Promethium into the XS-data sets in the PMAXS-format. The comparison of the results obtained with PARCS (using the corrected PMAXS file) and CASMO for the depletion problem exhibited a good agreement. Consequently, this approach was followed for the generation of a complete XS-set for a real BWR core to be used in subsequent transient analysis. Then 3D neutronic and thermal hydraulic core model were elaborated for a TRACE/PARCS analysis. The thermal hydraulic model is based on the 3D VESSEL

  15. Analysis of a German BWR core with TRACE/PARCS using different cross section sets

    International Nuclear Information System (INIS)

    'Full text:' Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools as well as appropriate cross sections (XS) libraries depending on the individual thermal hydraulic state parameters. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running and user-friendly lattice codes such as Casco and Helios. At research institutions and universities, alternative tools to the commercial ones with full access to the source code as well as moderate cost are urgently needed. The Scale system is being developed and improved for lattice physics calculations of real core loading of Light Water Reactors (LWR). It represents a promising alternative to the commercial lattice codes. At Karlsruhe Institute of Technology (Kit) a computational route based on Scale/Triton/Newt for BWR core loading is under development. The generated XS-data sets have to be transformed in PMAXS-format for use in the reactor dynamic code PARCS. This task is performed by the module GenPMAXS being developed and tested at the Michigan University. To verify the computational route, a BWR fuel assembly depletion problem was calculated by PARCS and compared to the CASMO results. Since the SCALE/TRITON XS-file does actually not contain all required neutronic data, FORTRAN routines have been developed to incorporate the missing data e.g. the yields of Iodine, Xenon and Promethium into the XS-data sets in the PMAXS-format. The comparison of the results obtained with PARCS (using the corrected PMAXS file) and CASMO for the depletion problem exhibited a good agreement. Consequently, this approach was followed for the generation of a complete XS-set for a real BWR core to be used in subsequent transient analysis. Then 3D neutronic and thermal hydraulic core model were elaborated for a TRACE/PARCS analysis. The thermal hydraulic model is based on the 3D VESSEL

  16. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  17. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR; BUTREN-RC un sistema hibrido para la optimizacion de recargas de combustible nuclear en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Castillo M, J.A. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  18. RELAP5/SCDAPSIM/MOD3.5 analysis of the influence of water addition during a core isolation event in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A.; Munshi, P. [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India)

    2014-07-01

    Highlights: • This work analyses the influence of water addition in a boiling water reactor during a Fukushima like scenario. • Injection uses thermal hydraulic conditions representative of the reactor core isolation cooling (RCIC) system. • A detailed RELAP/SCDAPSIM model of Laguna Verde BWR has been used for the analysis. • Results confirm the importance of both timing and temperature of core at the time of water addition to mitigate the accident. - Abstract: This work analyses the influence of water addition in a boiling water reactor during a core isolation event in BWR. Injection of water is impacted by time as well as reactor vessel water level and it uses thermal hydraulic conditions representative of the reactor core isolation cooling (RCIC) system. A detailed RELAP/SCDAPSIM model of Laguna Verde BWR vessel and related reactor cooling system (provided by the Mexican Nuclear Regulatory Authority) has been used for this analysis. These calculations have been extended to the point of likely vessel failure or stable core cooling. They focus on initial heating and melting of the core where water addition is found to be most effective in limiting the extent of fuel melting. It also presents the results of a base case, a station blackout transient without water addition. These calculations have been carried out up to 5 h (after reactor scram) beyond the point of likely vessel failure. The maximum core surface temperature of 3042 K and hydrogen production of 367 kg is observed in this case. The importance of timing can be seen from 3500 s and 3700 s injection cases. One case leads to maximum core surface temperature of 1520 K with hydrogen production of 21 kg while the second case leads to temperature of 2940 K with hydrogen production of 193 kg. Temperatures (at the time of start of first injection) in both these cases are1371 K and 1590 K which explains this switching from stable core cooling to very high core surface temperature.

  19. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user's manual

    International Nuclear Information System (INIS)

    This document is the User's Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code's capabilities and limitations; Chapter 2 describes the code's structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs

  20. Cooling system for reactor container

    International Nuclear Information System (INIS)

    Purpose: To effectively cool a reactor container upon reactor shutdown with no intrusion of metal corrosion products in coolants into the main steam pipe in a BWR type reactor. Constitution: A clean up system comprising a pipeway, a recycling pump, a non-regenerative heat exchanger and a primary coolant purifier and a regenerative heat exchanger is provided branched from a residual heat removing system and the clean up system is connected by way of a valve to a feedwater pipeway, as well as connected by way of the pipeway to the main steam pipeway at the midway of two main steam separation valves outside of the reactor container. This enables to prevent metal corrosion products floating on the surface of reactor water from introducing into the main steam pipe when the pressure vessel is filled with water. Then, since the pressure vessel is filled with primary coolants, the pressure vessel can be cooled uniformly in a short time. (Ikeda, J.)

  1. Investigation of void effects in boiling water reactor fuels using neutron tomography

    OpenAIRE

    Loberg, John

    2006-01-01

    In a boiling water reactor (BWR), the void is correlated to dry out and the power level of the reactor. However, measuring the void is very difficult so it is therefore calculated with an accuracy that leaves room for improvements. Typically the uncertainty is ± 3% for 40% average void in the reactor. If the void could be determined with improved accuracy, both safety and economical features could be improved. X-ray tomography has previously been done on BWR fuel models in order to determine ...

  2. Studies of a larger fuel bundle for the ABWR improved evolutionary reactor

    International Nuclear Information System (INIS)

    Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy

  3. Mobile crud and transportation of radioactivity in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H-P. [Studsvik Nuclear AB, Nykoping (Sweden); LTU, Div. of Chemical Engineering, Lulea (Sweden); Hagg, J. [Ringhals AB, Varobacka (Sweden)

    2010-07-01

    Mobile crud is here referred to as a generic term for all types of particles that occur in the reactor water in BWRs and that are able to carry radioactivity. Previous results in this on-going series of studies in Swedish BWRs suggest that there are particles of different origins and function. A share may come from fuel crud and others may come from detachment, precipitation and dissolution processes in different parts of the BWR primary system, as well as from other system parts, such as the turbine/condenser. In addition, crud particles in this sense may come from purely mechanical processes such as degradation of graphite containing parts of the control rod drives. Therefore, the overall aim was to evaluate which particles are responsible for the transportation and distribution of radioactivity and also to clarify the chemical conditions under which they are formed. Furthermore the aim was to draw conclusions about how the chemistry would be like in order to avoid or at least minimize the formation of radioactivity distributing particles. A specific objective has also been to look into the importance of particle size for spreading of radioactivity in the primary system. Different types of crud particles are likely to have different characteristics in terms of function associated with transportation of radioactivity. The fuel crud is radioactive from the source and other types of crud can via surface processes, co-precipitation and other chemical and mechanical processes potentially affect the distribution of radioactivity in the primary system. In order to predict how operating parameters (e.g. stable, full power operation and scram) and chemical parameters (NWC/HWC/Zn, etc.) will affect the activity build-up on the system surfaces, it is important to know how the different types of crud are affected by these and related parameters. Fuel crud fixed on cladding ring samples, as well as mobile crud from the reactor water captured on filters, were examined by

  4. 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

  5. Assessment of hydrogen combustion effects in the BWR/6 - Mark III Standard Plant

    International Nuclear Information System (INIS)

    This report discusses General Electric's study of potential hydrogen combustion effects on the Standard Mark III containment during postulated severe accidents. This study was performed as part of the Probabilistic Risk Assessment of the BWR/6 - Mark III Standard Plant. The methodology of determining the accident event sequence and modeling of the Boiling Water Reactor core response, including hydrogen generation by metal-water reaction, is described. Combustion of hydrogen released to the containment is analyzed and effects on the Mark III containment system are assessed. It is concluded that even for those cases where containment integrity may be lost, the containment function (i.e., limiting offsite doses) is maintained by the drywell and suppression pool

  6. Critical experiments for BWR fuel assemblies with cluster of gadolinia rods

    International Nuclear Information System (INIS)

    Gadolinia-bearing fuel rods are needed for high-burnup fuels. Strong neutron absorption of gadolinia makes an assembly heterogeneous from the viewpoint of reactor physics. The cluster of gadolinia-bearing fuel rods is useful for higher-burnup fuels than current fuels. Few critical experiments have been reported for fuel assemblies with the cluster of gadolinia-bearing fuel rods. We conducted critical experiments for BWR fuel assemblies with the cluster of gadolinia-bearing fuel rods in the Toshiba Nuclear Critical Assembly (NCA). Critical water level and power distribution were measured. Measurements were compared with analyses by a continuous-energy Monte Carlo code, MCNP, with the JENDL3.3 nuclear data library. (author)

  7. Study of transient rod extraction failure without RBM in a BWR

    International Nuclear Information System (INIS)

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

  8. Optimization of fuel reloads for a BWR using the ant colony system

    International Nuclear Information System (INIS)

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

  9. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  10. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms

    International Nuclear Information System (INIS)

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  11. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  12. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  13. Development of high performance catalyst for off-gas treatment system in BWR

    International Nuclear Information System (INIS)

    A high performance catalyst for off-gas treatment system in boiling water reactor (BWR) has been developed. The hydrogen concentration in the outlets of off-gas recombiners increased at several BWR plants in Japan. These phenomena were caused by deactivation of catalysts for the recombiners, and we assumed two types of deactivation mechanisms. The first cause was an increase of the amount of boehmite in the catalyst support due to alternation of the manufacturing process. The other cause was catalysts being poisoned by cyclic siloxanes that were introduced from the silicone sealant used in the upstream of the off-gas recombiners. The catalysts were manufactured by Pt adhering on alumina support. The conventional catalyst (CAT-A) used the aqueous solution of the chloroplatinic acid for adhesion of Pt. A dechlorination process by autoclave was applied to prevent the equipment at the downstream of the recombiners from stress corrosion cracking, but this process caused the support material to transform into boehmite. The boehmite-rich catalysts were deactivated more easily by organic silicon than gamma alumina-rich catalysts. Therefore, the CAT-A was replaced at many Japanese BWR plants by the improved catalyst (CAT-B), and their support was transformed into more stable gamma alumina by heating at 500degC. However, the siloxanes keep being detected in the off-gas though the source of siloxane had been removed and there still remain possibilities to deactivate the catalysts. Therefore, we have been developing high performance catalyst (CAT-C) that has higher activity and durability against poisoning. We investigated the properties of CAT-C by performance tests and instrumental analyses. The dependency of thermal output of nuclear reactor, and durability against siloxane poisoning were investigated. We found that CAT-C showed higher performance and better properties than CAT-B did. Moreover, we have been developing a modeling method to evaluate the hydrogen recombination

  14. 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.

  15. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  16. Experimental data report for test TS-5 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    This report presents experimental data for Test TS-5 which was the fifth test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in January, 1993. Test fuel rod used in the Test TS-5 was a short-sized BWR (7x7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79% and a burnup of 26GWd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The nominal energy deposition of 117±5cal/g·fuel (98±4cal/g·fuel in peak fuel enthalpy) was subjected to the test fuel rod and no fuel failure was observed in the test. The test fuel was pulse irradiated in a flow shroud which simulates fuel/water ratio in the commercial assembly. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

  17. 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

  18. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

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

  19. Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations

    International Nuclear Information System (INIS)

    The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

  20. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models

    International Nuclear Information System (INIS)

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  1. Modeling level instrumentation response to variations on drywell ambient conditions for a BWR

    International Nuclear Information System (INIS)

    For a BWR, the water level instrumentation is based on the differential pressure existing between two lines known as the reference and variable legs. The latter corresponds to the actual vessel level hydrostatic column from the reactor down to the differential pressure instrument location. The reference leg column is produced by piping connecting the instrumentation with a condensing chamber maintained at constant level. The calibration of the piping hydrostatic pressures associated to both legs is performed in BWRs using an average temperature representative of the drywell and one average temperature for the secondary containment. During reactor operation at steady state, there are temperature gradients in the ambient temperatures inside the primary containment that produce density changes in the water of both legs and will show a small change in the reactor level even when no physical change occurs. More important is the modification of the temperature distribution of the containment, which would lead to significant changes in the level reported by the instrumentation. In this work, a thermal model of the level instrumentation piping is developed to evaluate ambient temperature changes and gradients along the piping trajectory. The model takes into account axial and radial heat transfer for both the reference and variable legs. It is shown that possible changes in the drywell temperature may lead to apparent changes in level when no physical change occurs. In the practice, the model can be useful to evaluate the effect of heat and air conditioning systems inside the primary containment. (Author)

  2. BWR core thermal-hydraulic stability in anticipated transients without SCRAM

    International Nuclear Information System (INIS)

    The potential exists for core thermal hydraulic density wave oscillations to occur in the boiling water reactor (BWR) during anticipated transients without scram (ATWS) events. Conditions which may lead to oscillations in ATWS result either from recirculation pump trip (RPT) from appropriate ATWS signals, or from the failure to scram during core oscillations. A program to evaluate these scenarios will assess the adequacy of the system design and emergency operating procedures relative to the reactor system performance. Evaluation of ATWS events with oscillations is extremely complex, and defining the bases for the analyses is critical to the success of the program. Accurate analysis requires detailed modeling of the sequence of events, the plant configuration, and initial conditions. In conclusion: The likelihood and magnitude of any core thermal-hydraulic oscillations for a given ATWS event sequence depend not only on the choice of initial reactor conditions and fuel characteristics, but also on system dynamics and operator actions. Because of the complexity of the phenomena, accurate analysis requires detailed modeling of all relevant systems and interactions. In addition, selection of parameters must be consistent both with the intended application of the results and across the full spectrum of input conditions. Analysis of ATWS/stability events requires a reasonably limiting but representative set of plant conditions and accurate simulation of operator actions. (authors)

  3. 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

  4. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 12, 2011, 8:00 PM status

    International Nuclear Information System (INIS)

    This situation note is established according to the information gained on March 12, 2011, at 8:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2 and 3 of the Fukushima I site is briefly presented with the progress of the accident management actions. The operation principles of a BWR-type reactor and of a PWR-type reactor are presented in appendix as well as the confinement principle specific to Mark I-type BWR reactors designed by General Electric. The meteorological forecasts of the day are presented in a figure. (J.S.)

  5. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  6. 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)

  7. 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

  8. 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.

  9. Tendencias en la segregación residencial en Chile

    Directory of Open Access Journals (Sweden)

    Alan Garín Contreras

    2009-01-01

    Full Text Available Los análisis sobre segregación residencial en Chile, están adquiriendo una importancia creciente, dado que sus efectos sociales son cada vez más evidentes. A objeto de identificar las tendencias que sigue este proceso, se hace una revisión de la investigación a nivel nacional y que permita ser un aporte para el avance en este tema. Los resultados indican que la segregación residencial está disminuyendo, los principales mecanismo que operan en este proceso son: el funcionamiento del mercado del suelo, políticas públicas y los movimientos migratorios al interior de la ciudad.Comoefecto, la segregación ha generado un aumento de los problemas sociales como el alcoholismo, delincuencia, jóvenes solteras embarazadas y deserción escolar. También se observa una predominancia de estudios cuantitativos y una significativa heterogeneidad en el uso de indicadores, como un escaso análisis sobre la percepción y autopercepción de los habitantes de barrios segregados.

  10. 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)

  11. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  12. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  13. Influence of iron and nickel species upon activity buildup under simulated BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bjornsson, S.; Chen, J. [Studsvik Nuclear AB, Nykoping (Sweden); Lejon, J. [OKG AB, Oskarshamn (Sweden); Granath, G. [Ringhals AB, Varobacka (Sweden); Tanse-Larsson, M. [Forsmarks Kraftgrupp AB, Osthammar (Sweden)

    2010-07-01

    Activity build-up in BWR systems are of importance for service- and maintenance work performed at the plants. Minimizing the activity build-up is desirable for minimizing doses of personnel at the plants. Numerous studies have been carried out in this important field to understand the activity uptake mechanisms. This paper studied the possible role of Fe(II/III) and Ni(II) impurities in reactor water in activity uptake on stainless steel surfaces. The study was carried out by using a test loop with simulated BWR water containing Fe(II/III), Ni(II) and Co-60 marked Co(II) species of varied concentration and 500 ppb O{sub 2}. The test tube section in the loop system was pre-exposed type 316L stainless steel material. The microstructures of the formed oxide films were examined with high resolution electron microscopy (FE-SEM and FE-TEM). The activity monitoring on the test section showed that injection of 10 ppb Ni(II) and 0.1 ppb Fe(II/III) in the water with 0.1 ppb Co(II) was capable of stopping completely activity uptake. When Co(II) addition in the loop was stopped no activity return to the water could be seen. In another exposure test, injection of combined 2 ppb Fe(II/III) and 0.5∼10 ppb Ni(II) profoundly increased activity uptake on the test section with a maximum in activity buildup at 5 ppb Ni(II). When Co(II) addition in the loop was stopped a slight activity return was seen. The observed differences as seen in the two tests are discussed in view of the microstructures of the oxide films formed. (author)

  14. CSAU methodology and results for an ATWS event in a BWR using information theory methods

    International Nuclear Information System (INIS)

    Highlights: • We apply the CSAU methodology to an ATWS in a BWR using information theory methods. • We show how to perform the selection of the most influential inputs on the critical safety parameter. • We apply the maximum entropy principle to get the input parameter distribution. • We examine the maximum relative entropy principle to update the input parameter PDF. • We quantify the uncertainty of the critical safety parameter using order statistics and information theory. - Abstract: This paper shows an application of the CSAU methodology to an ATWS in a BWR reactor, when the temperature of the suppression pool is taken as the critical safety parameter. The method combines CSAU methodology with recent techniques of information theory. In this paper we use auxiliary tools to help in the evaluation and improvement of the parameters distribution that enter in the elements II and III of CSAU based methodologies. These tools have been implemented in two FORTRAN programs: GEDIPA (Generation of the Parameter Distribution) and UNTHERCO (Uncertainty in Thermal Hydraulic Codes). The first one analyzes the information data available on a given parameter or parameters with the goal to know all the information about the probability distribution function of these parameters. The second apply information theory methods, as the maximum entropy principle (MEP) and the maximum relative entropy Principle (MREP), in order to build conservative distribution functions for the parameters from the available data. Also, the distribution function of a given parameter can be updated using the MREP principle when new information is provided. UNTHERCO performs the MONTECARLO sampling for a given set of parameters when the distribution function of these parameters is previously known. If the distribution of a parameter is unknown, then, the MEP is applied to deduce the distribution function for this parameter

  15. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  16. Hydrogen uptake of BWR fuel rods. Power history effects at long irradiation times

    International Nuclear Information System (INIS)

    AREVA LTP (Low Temperature Process) Zircaloy-2 cladding for Boiling Water Reactors (BWR) in both RXA (Recrystallized Annealed) and CWSR (Cold Worked Stress Relieved) metallurgical states, has an optimized microstructure with an optimum size of SPP (Secondary Phase Particles) that has reduced the nodular corrosion to a minimum while maintaining a good uniform corrosion performance with acceptable hydrogen pickup. Classically hydrogen uptake is described by the Hydrogen Pick-Up Fraction (HPUF), which is the ratio of the hydrogen generated by uniform oxidation that is eventually picked up by the metal to the total hydrogen generated by oxidation. In the past, the hydrogen uptake database showed a low HPUF with hydrogen concentration close to the saturation value of the metal at operating temperature and correspondingly little hydride formation. The hydrogen concentration was correlated with irradiation time via the HPUF (at an almost constant corrosion and hydrogen production rate). Recently, some significantly higher hydrogen concentration values (300 wppm and more) have been measured for medium and high burnup rods. This effect was also observed on four AREVA fuel rods from BWR (Boiling Water Reactors). This prompted a thorough analysis of the hydrogen pickup database as well as material and environmental factors influencing corrosion and hydrogen uptake. The most important outcome of the investigation was that a low power – low steam condition is associated with increased hydrogen pickup. The linear power is a proxy variable for low heat flux and low steam quality in the coolant, which were identified as important parameters for physical processes that could explain the enhanced hydrogen uptake in some cases. The paper will present the database of the enhanced hydrogen uptake measured in European power reactors and demonstrate the effect of power history on the uptake process. Power histories with high hydrogen uptake included extended low power periods later in

  17. IASCC susceptibility under BWR conditions of welded 304 and 347 stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Castano, M.L. [CIEMAT, Complutense 22, 28040 Madrid (Spain); Schaaf, B. van der [NRG, Petten (Netherlands); Roth, A. [Framatome ANP, Erlangen (Germany); Ohms, C. [JRC-IE, Petten (Netherlands); Gavillet, D. [PSI, Villigen (Switzerland); Dyck, S. van [SCK - CEN, Mol (Belgium)

    2004-07-01

    In-service cracking of Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) internal components has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC), a high temperature degradation process that austenitic stainless steels exhibit, when subjected to stress and exposed to relatively high fast neutron flux. Most of the cracking incidents in BWRs were associated to the heat-affected zone (HAZ) of welds. Although the maximum end-of- life dose for this structure is about 3 x 10{sup 20} n/cm{sup 2}, below the threshold fluence of 5 x 10{sup 20} n/cm{sup 2} (equivalent to {approx} 1 dpa) for IASCC in BWR of annealed materials, the influence of neutron irradiation in the weld and HAZ is still an open question. As a consequence of the welding process, residual stresses, microstructural and microchemical modifications are expected. In addition, exposure to neutron irradiation can induce variations in the material's characteristics that can modify the stress corrosion resistance of the welded components. While the IASCC susceptibility of base materials is being widely studied in many international projects, the specific conditions of irradiated weldments are rarely assessed. The INTERWELD project, partially financed by the 5. Framework program of the European Commission, was defined to elucidate neutron radiation induced changes in the HAZ of austenitic stainless steel welds that may promote intergranular cracking. To achieve this goal the evolution of residual stresses, microstructure, micro-chemistry, mechanical properties and the stress corrosion behaviour of irradiated materials are being evaluated. Fabrication of appropriate welds of 304 and 347 stainless steels, representative of core components, was performed. These weld materials were irradiated in the High Flux Reactor (HFR) in Petten to two neutron dose levels, i.e. 0.3 and 1 dpa. Complete characterization of the HAZ of both materials, before and after irradiation is

  18. BWR core stability analysis with RETRAN

    International Nuclear Information System (INIS)

    A RETRAN model was developed for determining the stability of boiling water reactor. This model was benchmarked against plant data from stability tests conducted during plant operations. The stability analysis with RETRAN is demonstrated using best estimate RETRAN input data representative of the NSSS. All of the important neutronic and thermal hydraulic feedback mechanisms are taken into account through the modeling of the reactor vessel, recirculation loops and core neutronics. The analysis was performed with the RETRAN02 MOD003 computer code. The transient is initialized by a small step decrease in the steam dome pressure. The core exit (upper plenum) pressure and core power transient responses to this perturbation are transformed into frequency data and a system transfer function is then obtained. The system transfer function is fitted to a second order equation from which the decay ratio and natural frequency can be determined

  19. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

  20. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE

  1. BWR refill-reflood program: core spray distribution experimental task plan

    Energy Technology Data Exchange (ETDEWEB)

    Eckert, T.

    1981-02-01

    An experimental task plan for the BWR/4 core spray task of the Refill-Reflood Test Program is presented. The test program will provide core spray distribution data for a 30 degree sector of the BWR/4 and 5-218 design. This design uses different nozzle types and different sparger elevations than the BWR/6-218 design which was tested previously. Test parameter ranges are specified; individual tests are defined; and measurement and data utilization plans are defined.

  2. BWR drywell behavior under steam blowdown

    International Nuclear Information System (INIS)

    Historically, thermal hydraulics analyses on Large Break Loss of Coolant Accidents (LOCA) have been focused on the transients within the reactor or steam generator. Few have studied the effects of steam blowdown on the containment building. This paper discusses some theoretical issues as well as presenting numerical and experimental results of the blowdown tests performed at the Purdue University Multi-Dimensional Integrated Test Assembly (PUMA)

  3. Application of water jet penning technology to BWR core shroud for IGSCC mitigation

    International Nuclear Information System (INIS)

    Water Jet Peening (WJP) is one of the promising SCC mitigation technologies which make original surface tensile residual stress to compressive one. The Water Jet Peening Technology has the following advantages: a) no foreign material entering into the reactor because of using only water, b) applicability to narrow and complicated structure because it is effective in the wide range of parameters, c) simple in the system/equipment and short period of application in actual plant. WJP was first applied to BWR Core Shroud for preventive maintenance purpose during 1999 outage in Japan. Although the target welds of Shroud are surrounded by various kinds of other components and access space is very limited, most of the weld could be peened by optimizing the peening condition. Effect of residual stress improvement was verified by mock-up test prior to actual work. WJP application was completed within the planned schedule without trouble. Application experience to the Shroud and examples of development of application to other Reactor Internal components will be presented. (author)

  4. Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

  5. An overview of third generation reactors - Presentation

    International Nuclear Information System (INIS)

    Now, in 2013 about 69 nuclear reactors (67 GWe) are being built in the world, mostly in China and Russia. Although a few second generation reactors are being built (particularly in China), third generation reactors seem to be systematically chosen for any new construction. The French commercial offer is based on 3 models: the EPR, the ATMEA-1 and the KERENA, the first 2 being of PWR-type while the latter is a BWR. A lot of third generation reactor designs are available on the international market. Concerning the PWR technology we have: the EPR (AREVA), the AP1000 (Westinghouse), the AES 2006 and the VVER TOI (Rosatom), the APR1400 (KEPCO), the APWR (MHI), the ATMEA-1 (AREVA + MHI), the ACP 1000 (CNNC - China), ACPR1000 (CGN - China) and CAP 1400 (SNPTC - China). Concerning the BWR technology, the commercial offer is far less important we have: the ABWR (GE-Hitachi + Toshiba) and the ESBWR (GE-Hitachi). Some third generation reactors are operating now: the first ABWR and AES reactors but most of them (EPR, AP1000, APR1400 and AES 2006) are at an advanced stage of construction. (A.C.)

  6. BWR shutdown and startup chemistry experience and application Sourcebook. BWRVIP-225, Rev. 1

    International Nuclear Information System (INIS)

    BWR water chemistry has changed significantly over the years with the adoption of hydrogen water chemistry (HWC), noble metal chemical application (NMCA), and most recently, Online NobleChem™ (OLNC). Some plants have experienced large increases in activated corrosion products during shutdown evolutions, when the chemistry environment at primary system surfaces transitions from reducing to oxidizing conditions. Higher activity releases may be in part related to the more reducing conditions brought about by the above mentioned processes during the operating cycle. With shorter outages decreasing the available cleanup time, some plants are experiencing increased outage radiation exposure. A significant portion of fuel cycle intergranular stress corrosion cracking (IGSCC) propagation of reactor internals and primary system piping is indicated to occur during startup and early power ascension, when dissolved oxygen and hydrogen peroxide concentrations in the reactor coolant are high and hydrogen injection is unavailable. The majority of lost hydrogen availability hours typically occurs during early startup. Startup periods following refueling outages are also when reactor coolant chemistry transients may occur due to system flow changes and residual chemical impurities from outage related work activities. Test results show that IGSCC is accelerated particularly during early startup periods of elevated reactor coolant oxidant concentrations (dissolved oxygen and hydrogen peroxide), particularly when operating at an intermediate temperature range (300 – 400°F, 148-204°C). Based on extensive data collection and evaluation, BWRVIP-225 Revision 1 provides good practices and conditions to avoid during plant refueling outages, including recommendations to minimize activity transport during shutdown conditions to reduce radiation exposure. In addition, good practices and conditions to avoid are provided for startup and power ascension to minimize IGSCC. This paper

  7. Optimization of fuel cells for BWR based in Tabu modified search; Optimizacion de celdas de combustible para BWR basada en busqueda Tabu modificada

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Palomera P, M.A. [Facultad de Ingenieria, UNAM, Posgrado en Ingenieria en Computacion, Circuito exterior s/n, Ciudad Universitaria, Mexico, D.F. (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2004-07-01

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

  8. Design and optimization of a fuel reload of BWR with plutonium and minor actinides; Diseno y optimizacion de una recarga de combustible de BWR con plutonio y actinidos menores

    Energy Technology Data Exchange (ETDEWEB)

    Guzman A, J. R.; Francois L, J. L.; Martin del Campo M, C.; Palomera P, M. A. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: maestro_juan_rafael@hotmail.com

    2008-07-01

    In this work is designed and optimized a pattern of fuel reload of a boiling water reactor (BWR), whose fuel is compound of uranium coming from the enrichment lines, plutonium and minor actinides (neptunium, americium, curium); obtained of the spent fuel recycling of reactors type BWR. This work is divided in two stages: in the first stage a reload pattern designs with and equilibrium cycle is reached, where the reload lot is invariant cycle to cycle. This reload pattern is gotten adjusting the plutonium content of the assembly for to reach the length of the wished cycle. Furthermore, it is necessary to increase the concentration of boron-10 in the control rods and to introduce gadolinium in some fuel rods of the assembly, in order to satisfy the margin approach of out. Some reactor parameters are presented: the axial profile of power average of the reactor core, and the axial and radial distribution of the fraction of holes, for the one reload pattern in balance. For the design of reload pattern codes HELIOS and CM-PRESTO are used. In the second stage an optimization technique based on genetic algorithms is used, along with certain obtained heuristic rules of the engineer experience, with the intention of optimizing the reload pattern obtained in the first stage. The objective function looks for to maximize the length of the reactor cycle, at the same time as that they are satisfied their limits related to the power and the reactor reactivity. Certain heuristic rules are applied in order to satisfy the recommendations of the fuel management: the strategy of the control cells core, the strategy of reload pattern of low leakage, and the symmetry of a quarter of nucleus. For the evaluation of the parameters that take part in the objective function it simulates the reactor using code CM-PRESTO. Using the technique of optimization of the genetic algorithms an energy of the cycle of 10834.5 MW d/tHM is obtained, which represents 5.5% of extra energy with respect to the

  9. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network

    International Nuclear Information System (INIS)

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U235, some of these bars also contain a concentration of Gd2O3 and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  10. Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

  11. Development of evaluation tool for radiation dose rate distribution in PCV of Hamaoka BWR plants based on water chemistry

    International Nuclear Information System (INIS)

    We have developed an evaluation tool for the radiation dose rate distribution of the work areas in the primary containment vessel (PCV) of Units 3, 4 (BWR5) and 5 (ABWR) at Hamaoka NPS. This tool has been constructed based on the transport behavior of radioactive corrosion products in the primary cooling water of BWR. This tool can be used to evaluate quantitatively the effects of the dose reduction methods by water chemistry control or radiation management. It is composed of two calculation codes; water chemistry code (ACTTUBE) and radiation dose rate code (RADTUBE). ACTTUBE calculates the piping dose rates based on the mass balance of corrosion products, 6 kinds of metal and 5 kinds of radionuclide, among the parts of primary cooling water, such as reactor water, feed water, fuel rod surface and out-of-core piping surface. RADTUBE calculates the dose rate distribution based on the radiation shielding calculation from a calculation result of ACTTUBE. Additionally, this tool has a visualization function of calculated radiation dose rate distribution in the PCV by using a wireless controller and 3D glasses/monitor in order to improve user convenience. The accuracy of the tool's calculation results was evaluated using the water chemistry data and radiation dose rate data of the Hamaoka plants. As a result, it was confirmed that this tool had sufficient accuracy to be used in the evaluation of radiation dose rates for the radiation management of actual plants. (author)

  12. 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.)

  13. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  14. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, Carl, E-mail: carl.adamsson@psi.ch [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden); Le Corre, Jean-Marie, E-mail: lecorrjm@westinghouse.com [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden)

    2011-08-15

    Highlights: > The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. > A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. > MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. > The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. > The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle

  15. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  16. LAPUR-K BWR stability benchmark

    International Nuclear Information System (INIS)

    This paper documents the stability benchmark of the LAPUR-K code using the measurements taken at the Ringhals Unit 1 plant over four cycles of operation. This benchmark was undertaken to demonstrate the ability of LAPUR-K to calculate the decay ratios for both core-wide and regional mode oscillations. This benchmark contributes significantly to assuring that LAPUR-K can be used to define the exclusion region for the Monticello Plant in response to recent US Nuclear Regulatory Commission notices concerning oscillation observed at Boiling Water Reactor plants. Stability is part of Northern States Power Reload Safety Evaluation of the Monticello Plant

  17. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  18. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H. P

    2002-03-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

  19. Inservice inspection of Halden BWR pressure vessel

    International Nuclear Information System (INIS)

    A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

  20. Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

    International Nuclear Information System (INIS)

    The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

  1. Concept of the core for a small-to-medium-sized BWR that does not use control rods during normal operation

    International Nuclear Information System (INIS)

    A small-to-medium-sized boiling water reactor (BWR) with a natural circulation system is being developed for countries where initial investment funds for construction are limited and electricity transmission networks have not been fully constructed. To lighten operators' work load, a core that does not use control rods during normal operation (control rod-free core) was developed by using a neutronics calculation system coupled with core flow evaluation. The control rod-free core had large core power fluctuation with conventional burnable poison design. The target of core power fluctuation was set to less than 10% and was achieved by optimization of burnable poison arrangement. (author)

  2. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

  3. Estudios sobre educación intercultural en Colombia: tendencias y perspectivas

    Directory of Open Access Journals (Sweden)

    Celmira Castro Suárez

    2009-01-01

    Full Text Available En este artículo la autora da cuanta de las tendencias y perspectivas de los estudios sobre educación intercultural que se han realizado en el país en los últimos años. Para lo cual hace un recorrido por los diversas preocupaciones por parte de los estudiosos del tema, quienes desde distintas tendencias conceptuales y disciplinares han venido redefiniendo el objeto de estudio de la diversidad cultural, la cual ya no solo es abordada con elpropósito de cuantificar etnias o grupos culturales existentes, o para relacionar e inventariar sus lenguas o identificar las tendencias pedagógicas de moda en relación a lacultura.

  4. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms; Una metodologia para obtener los patrones de barras de control en un BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Montes T, J.L. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Requena R, I. [Universidad de Granada, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

    2003-07-01

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  5. Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

  6. Ten year's experience of in-service inspection on BWR vessel stub tubes

    International Nuclear Information System (INIS)

    The stub tube is a component of the control rod housings in boiling water reactor (BWR) nuclear power plants. In certain cases these tubes may undergo cracking, as a result of which fluid leakage may occur from the reactor vessel. Consequently, these components have to be inspected during service in order to determine whether or not they are affected by such defects. The stainless steel/Inconel stub tubes are welded at the upper end to the control rod housing, and at the lower end to the reactor vessel. Given the geometry, material, welds, stresses and corrosive elements associated with these components, intergranular corrosion cracking may occur in the areas adjacent to the welds. For this reason inspections capable of detecting this type of defect must be performed, with a view to determining the integrity of the component. Since 1981, more than 300 stub tube inspections have been carried out at different Spanish nuclear power plants. Initially, a single ultrasonic technique was used to detect the presence of indications; at present, and after several intermediate stages, various ultrasonic and eddy current techniques are used to dimension the length and depth of indications, determine their evolution and ensure dimensional control of the component for subsequent repair. Parallel to the development of non destructive testing techniques, mechanical scanning equipment has been designed and manufactured for use in test performance. Throughout development of these techniques, and prior to application in the field, different validation tests have been performed, initially using blocks containing artificial reflectors and subsequently blocks with actual crack-type reflectors. (author)

  7. Effect of UV irradiation on low concentration methanol solutions in BWR condition loop testing

    International Nuclear Information System (INIS)

    The reactor pressure vessel (RPV) internals play a significant role in BWRs with respect to ensuring the function of several neutron flux controlling components in the reactor core. Effective countermeasures to prevent the RPV internals from stress corrosion cracking (SCC) are needed, especially, if locally sensitized or cold-worked materials are exposed to oxygenated high-temperature water (HTW). Consequently, mitigation techniques are necessary to reduce the dissolved oxidant concentration as oxygen and hydrogen peroxide to shift the corrosion and redox potentials of materials to more negative values (as e.g. ECP < -230 mVSHE for austenitic stainless steel). The beneficial effect of the alternative reductant MeOH was confirmed by test runs in a pipe reactor especially designed for these tests at T = 150 deg. C. One of the most important results is that MeOH injection in oxygenated high-temperature water (HTW) with simultaneous irradiation of the test solution by Vacuum UV (VUV) light (photolysis) is sufficient to significantly shift the corrosion potentials to more negative values already at a molar ratio (MeOH/DO) of ≥ 1. The photon energies of VUV light (6.2 to 12.4 eV) are high enough to crack the bonds in the water molecule and to generate radicals, ions and free electrons. VUV-light, thus, generates effects similar to Cherenkov radiation with a wavelength of 100 to 400 nm in the vessel of LWRs. The early injection techniques during plant start-up may become one of the advantages of MeOH injection at BWR plants. Hence, the MeOH-effect at molar ratios between 1 and 2 has to be studied more carefully to be sure that in this temperature regime the dosage of MeOH is reasonable. (authors)

  8. 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)

  9. Assessment with coupled thermo-mechanical creep analysis of combined CRGT and external vessel cooling efficiency for a BWR

    International Nuclear Information System (INIS)

    In this paper we consider in-vessel stage of a severe core melt accident in a Nordic design Boiling Water Reactor (BWR). Decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. Performed thermo-mechanical creep analysis identified two different modes of vessel wall failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Next, given the mechanical and thermal loads from the decay-heated melt, external vessel cooling is applied at a specified time. It is found that combined CRGT and external vessel cooling was able to suppress the creep and subsequently prevent vessel wall failure. (author)

  10. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  11. IAEA activities in nuclear reactor simulation for educational purposes

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Currently, the IAEA has simulation programs available for distribution that simulate the behaviour of BWR, PWR and HWR reactor types. (authors)

  12. Reliability improvement method for BWR power plants

    International Nuclear Information System (INIS)

    The construction of the nuclear power generating facilities in Japan was commenced by the import of technological know-how from the United States, but in ten years since then, they reached the stage of improvement and standardization by the effort for the domestic production and the accumulated technological ability. But the unscheduled stop of operation was not able to avoid centering around the initially imported plants, and it cannot be said that the sufficient rate of operation was attained. In Japan, plant manufacturers deliver the whole installations including nuclear reactors in the lump, and carry out the planning, design, manufacture, construction, periodic inspection and maintenance, accordingly the feedback of the operational results can be made quickly, differing from the U.S. system. As the result, No. 1 plant of the Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc., which was constructed by the domestic technology, has attained about 72% of the average rate of operation in six years, and showed the high reliability of the domestically produced plants. The measures for improving system reliability in system planning and the reliability of machinery and equipments, the method of evaluating the reliability of systems, machines and equipments, the quality of nuclear power generating facilities and the quality assurance, and the management of maintenance in Hitachi Ltd. are explained. (Kako, I.)

  13. A BWR power plant simulator for Barsebaeck

    International Nuclear Information System (INIS)

    A computer simulator of a Barsebaeck power plant unit has been developed in cooperation between Sydkraft AB, Lund Institute of Technology, and Risoe National Laboratory. The simulator is of the kind often referred to as a compact simulator, because it involves only a computer with display screens and other input/output devices plus the software needed for calculation and presentation of the plant state as a function of time, and no sort of model of the control room as in large reactor simulators for operator training. The purpose of training courses with the compact simulator is to give students a better understanding of the behaviour of the power plant under transient conditions by displaying variables, e.g. pressures, temperatures, reactivity, nuclear power, as functions of time, thereby showing the interactions between different parts of the plant during the transient and the influence of a number of possible operator actions. The present paper describes the Barsebaeck compact simulator with the emphasis on the software developed at Risoe National Laboratory. The Risoe work comprises the programming of the dynamic plant model, in the form of a number of Fortran subroutines containing the physical description of the power plant. (author)

  14. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  15. OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    The OECD Halden Reactor project is an agreement between OECD member countries. It was first signed in 1958 and since then regularly renewed every third year. The activities at the Project is centred around the Halden heavy water rector, the HBWR. The reseach programme comprizes studies of fuel performance under various operating conditions, and the application of computers for process control. The HBWR is equipped for exposing fuel rods to temperatures and pressures, and at heat ratings met in modern BWR's and PWR's. A range of in-core instruments are available, permitting detailed measurements of the reactions of the fuel, including mechanical deformations, thermal behaviour, fission gas release, and corrosion. In the area of computer application, the studies of the communication between operator and process, and the surveillance and control of the reactor core, are of particular interst for reactor operation. 1988 represents the 30th year since the Project was started, and this publication is produced to mark this event. It gives and account of the activities and achievements of the Project through the years 1958-1988

  16. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Andrey [Paul Scherrer Institut, Villigen (Switzerland); Degueldre, Claude, E-mail: claude.degueldre@psi.ch [Paul Scherrer Institut, Villigen (Switzerland); Kaufmann, Wilfried [Kernkraftwerk Leibstadt, Leibstadt (Switzerland)

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  17. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  18. IASCC crack growth rate of neutron irradiated low carbon austenitic stainless steels in simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Chatani, K. [Nippon Nuclear Fuel Development Co. Ltd (NFD), Oarai (Japan); Takakura, K.; Ando, M.; Nakata, K. [Japan Nuclear Energy Safety Organization (JNES), Tokyo (Japan); Tanaka, S. [Toshiba Corp., Yokohama (Japan); Ishiyama, Y. [Hitachi Ltd., Hitachi (Japan); Hishida, M. [Inst. of Research and Innovation (IRI), Tokyo (Japan); Kaji, Y. [Japan Atomic Energy Agency (JAEA), Tokai (Japan)

    2007-07-01

    Crack Growth Rate (CGR) tests have been conducted with neutron irradiated Compact Tension (CT) specimens. The specimens were irradiated at core region of Japan Material Testing Reactor (JMTR) in simulated BWR water environments at 288 {sup o}C. The CGR tests of 316L and 304L base metals irradiated from 0.516 to 1.07 x 10{sup 25} n/m{sup 2} (E>1MeV), 316L and 308L weld metals irradiated up to 0.523 to 0.541 x 10{sup 25} n/m{sup 2} (E>1MeV) were performed with reversing DC potential drop method under constant load in a few stress intensity factor (K) and corrosion potential (ECP) conditions at 288 {sup o}C in water. CGRs of base metals were increased with increasing neutron fluence, Clear reductions in CGRs of base metals and weld metals were measured with decreasing ECP levels. This paper will discuss the relationship between CGR and radiation hardening / RIS. (author)

  19. Performance analysis of passively safe BWR with experimental and numerical simulation

    International Nuclear Information System (INIS)

    The performance of passive safety systems of a natural circulation BWR in a Large Break Loss Of Coolant Accident (LB LOCA) is evaluated with integral tests using a scaled test facility and RELAP5 (Mod3.3) code simulation. The Main Steam Line Break (MSLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) with the initial conditions given by the code simulation. The PUMA facility is designed to reproduce thermal-hydraulic phenomena during the low-pressure blowndown and long-term cooling period of the LOCA transient. The MSLB test is initialized when Reactor Pressure Vessel (RPV) depressurizes to 1 MPa (150 psi) and lasts for 8 hours. This test aims to demonstrate the performance of passive safety systems during the LB LOCA. Test results show that core heat-up is not observed during the test transient due to the function of Emergency Core Cooling System (ECCS). The containment peak pressure and temperature are below the design limit, which is mainly contributed by the function of Passive Containment Cooling System (PCCS). The MSLB accident transient has been simulated with RELAP5 code using prototypic plant mode and test facility model. The code models give reasonably accurate predictions on most system behaviors, while having some distortions for certain local phenomena. The integral test scalability and code applicability are evaluated by comparing the test data and the code simulation results, taking into consideration of the scaling methodology and code uncertainties. (author)

  20. Optimization of fuel cells for BWR based in Tabu modified search

    International Nuclear Information System (INIS)

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

  1. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  2. BWR core stability prediction on-line with the computer code matstab

    International Nuclear Information System (INIS)

    MATSTAB is a computer program for three-dimensional prediction of BWR core stability in the frequency domain. This tool has been developed, and is currently used, to perform core design and optimisation with regard to core stability. The requirement regarding the predicted decay ratio of the new core is one of the limiting factors, or key parameters, in core design. To be useful, the tool should be fast and simple to apply. The results must be delivered promptly and experts should not be required to interpret them. Alternatively, the area of application for MATSTAB can be described as on-line monitoring using predictive tools. Core stability properties can be calculated for a number of presumptive reactor states, planned or unplanned. A 3-D code operating in the frequency domain may be the best tool to use for the purposes just mentioned. Some strong advantages are that the results are given promptly, they require no post-processing and are directly amenable to graphic presentation of eigenvectors, etc. (authors)

  3. Influence of metal addition to BWR water on contamination and corrosion of stainless steel

    International Nuclear Information System (INIS)

    Oxide layers grown on stainless steel under modified BWR conditions with or without addition of different bivalent metal ions have been characterised using methods like SIMS and photo-electrochemistry. The Co-58 activity of the samples depends strongly on the thickness of the oxide film. Low pHT values generally favour dominance of p-type semiconductivity, implying a corrosion process controlled by cation transport through the oxide layer. High pHt values normally result in a change of semiconducting properties from predominantly p-type to n-type, which can be used as an indicator for a change in the corrosion mechanism. The metal ions added to high-temperature water do not significantly affect the specific activity of the different stainless steel samples after exposure. The aim of the described tests was to identify possible alternatives to zinc and to elucidate underlying mechanisms controlling the incorporation of radio-isotopes of cobalt in the oxide layer on stainless steel. Manganese has been identified in the described short-term exposure tests as a possible alternative to zinc. The positive effect of manganese must be confirmed by long-term tests. During these tests, the concentration of dissolved manganese should be reduced to 10 ppb maximum and the other water chemistry parameters should be also adjusted more closely to reactor coolant conditions. (orig./MM)

  4. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

    1999-07-01

    This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

  5. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  6. 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.

  7. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  8. Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

  9. 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

  10. 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.

  11. 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...

  12. 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

  13. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    International Nuclear Information System (INIS)

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

  14. Advanced ultrasonic inspection system for the ID-inspection of reactor pressure vessels of BWRs

    International Nuclear Information System (INIS)

    A newly-developed, modular ultrasonic examination system has been developed by Siemens for the ID inspection of BWR RPV'S. It is based on the phased-array technique with hybrid probes using the latest in manipulator and control equipment technology to allow the often hard-to-access weld areas of older reactor pressure vessels in US BWR plants to be examined within a very short time and with minimal radiation exposure of the examination personnel. New NRC stipulations requiring almost complete ultrasonic examination of all RPV welds can be fully satisfied using this system for the ID inspection of all longitudinal and circumferential welds above the jet pump baffle plate

  15. Severe accident assessment. Results of the reactor safety research project VAHTI

    International Nuclear Information System (INIS)

    The report provides a summary of the publicly funded nuclear reactor safety research project Severe Accident Management (VAHTI). The project has been conducted at the Technical Research Centre of Finland (VTT) during the years 1994-96. The main objective was to assist the severe accident management programmes of the Finnish nuclear power plants. The project was divided into five work packages: (1) thermal hydraulic validation of the APROS code, (2) core melt progression within a BWR pressure vessel, (3) failure mode of the BWR pressure vessel, (4) Aerosol behaviour experiments, and (5) development of a computerized severe accident training tool

  16. Effects of Void Uncertainties on Pin Power Distributions and the Void Reactivity Coefficient for a 10X10 BWR Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.; Krouthen, J.; Helmersson, S.; Chawla, R

    2004-03-01

    A significant source of uncertainty in Boiling Water Reactor physics is associated with the precise characterisation of the axially-dependent neutron moderation properties of the coolant inside the fuel assembly channel, and the corresponding effects on reactor physics parameters such as the lattice neutron multiplication, the neutron migration length, and the pin-by-pin power distribution. In this paper, the effects of particularly relevant void fraction uncertainties on reactor physics parameters have been studied for a BWR assembly of type Westinghouse SVEA-96 using the CASMO-4, HELIOS/PRESTO-2 and MCNP4C codes. The SVEA-96 geometry is characterised by the sub-division of the assembly into four different sub-bundles by means of an inner bypass with a cruciform shape. The study has covered the following issues: (a) the effects of different cross-section data libraries on the void coefficient of reactivity, for a wide range of void fractions; (b) the effects due to a heterogeneous vs. homogeneous void distribution inside the sub-bundles; and (c) the consequences of partly inserted absorber blades producing different void fractions in different sub-bundles. (author)

  17. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  18. Simplified system for the pressure control of a Nucleo electric central of the BWR type

    International Nuclear Information System (INIS)

    One of the main preoccupations of the electric power generator stations is the appropriate operation of the same ones. The operators must be qualified to respond in an adequate way and to be able to take to these power stations to an optimal, sure and stable operation condition under any circumstance. The Laboratory of Analysis in Nuclear Reactors Engineering (LAIRN) of the Engineering Faculty of UNAM (Fl) in collaboration with the International Atomic Energy Agency (IAEA), it develops an interactive classroom simulator in which simulations of the phenomena which take place in a nuclear power station are executed. The classroom simulator bases its operation on specialized nuclear codes feeding interactive graphic unfolding with those that it is possible to make a monitoring, supervision and control of the behavior of the power station under any operation regime, either in normal operation, transitory events or postulated accident sequence. The development of this classroom simulator includes a modular and re configurable structure. Due to it is indispensable to count with a higher inter activity with the system it is included the simulation of the control system of the plant and inside the same, one of those more important it is the reactor pressure control system. The present work describes the conceptual design and the used methodology for the development and implementation in the simulator of a simplified model of the pressure control system for a BWR generic central. The reach of the development will allow to accomplish the necessary tests to demonstrate that this has an adequate performance according to the carried out simplifications. (Author)

  19. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  20. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    OpenAIRE

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use ...

  1. 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.

  2. 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)

  3. 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

  4. Behavior of BWR-type fuel elements with B{sub 4}C/steel absorber tested under severe fuel damage conditions in the CORA facility

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L.; Hagen, S.; Hofmann, P.; Schanz, G.

    2009-01-15

    The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program was to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations were tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The BWR-type bundles consisted of 18 fuel rod simulators (heated and unheated rods), an absorber blade of steel containing eleven absorber rods filled with boron carbide powder. The larger bundle CORA-18 contained the same number of absorber rods but was made up of 48 fuel rod simulators. All BWR bundles were surrounded by a zircaloy shroud and the absorber blades by a channel box wall on each side, also made of zircaloy. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence was simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. In all experiments bundle destruction started in the upper region (axially) with melting of the absorber blade and the absorber rod cladding at about 1250 C by interaction of boron carbide and steel. After destruction of the channel box walls this melt attacked the zircaloy fuel rod cladding and started to interact with the UO{sub 2} pellets. The test bundles also resulted in severe oxidation of the following components made of zircaloy: shroud, cladding, and grid spacers at the central and upper positions. Relocated absorber melt

  5. Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

    International Nuclear Information System (INIS)

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

  6. Hydrogen generation from aluminium corrosion in reactor containment spray solutions

    International Nuclear Information System (INIS)

    The aluminium corrosion experiments in reactor containment spray solutions, under the conditions expected to prevail during LOCA in BWR and PWR, were performed in order to investigate relationships between temperature, pH and hydrogen production rates. In order to simulate the conditions in a BWR containment realistic ratios between aluminium surface and water volume and between aluminium surface and oxygen volume were used. Three different aluminium alloys were exposed to spray solutions: AA 1050, AA 5052 and AA 6082. The corrosion rates were measured for BWR solutions (deaerated and aerated) with pH 5 and 9 at 50, 100 and 1500C. The pressure was constantly 0.8 MPa. The hydrogen production rate was measured by means of gas chromatography. In deionized BWR water the corrosion rates did not exceed about 0.05 mm/year in all cases, i.e. were practically independent of temperature and pH. Hydrogen concentrations were less than 0.1 vol.% in cooled dry gas. Corrosion rates and hydrogen production in PWR alkaline solution measured at pH 9.7 and 1500C were very high. AA 5052 alloy was the best material

  7. 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)

  8. Aportación a los Cálculos Neutrónicos y Termohidráulicos en 3D con Códigos de Mejor EStimación. Aplicación a Transitorios en Reactores Nucleares BWR y PWR

    OpenAIRE

    SÁNCHEZ HERNÁNDEZ, ANA MARÍA

    2012-01-01

    El uso de códigos es una herramienta fundamental en Seguridad Nuclear para la simulación de diferentes situaciones en reactores de potencia. En particular, los códigos termohidráulicos de estimación óptima nos permiten simular de forma más realista los fenómenos que suceden en una central nuclear con la representación del circuito primario. A su vez los códigos neutrónicos de dinámica del núcleo, nos permiten una definición y simulación más precisa del núcleo. El uso de los códigos acopl...

  9. Huitzoctli: A system to design Control Rod Pattern for BWR's using a hybrid method

    International Nuclear Information System (INIS)

    Highlights: → The system was developed to design Control Rod Patterns for Boiling Water Reactors. → The critical reactor core and the thermal limits were fulfilled in all tested cases. → The Fuel Loading Pattern remains without changes during the iterative process. → The system uses the heuristics techniques: Scatter Search and Tabu Search. → The effective multiplication factor keff at the EOC was improved in all tested cases. - Abstract: Huitzoctli system was developed to design Control Rod Patterns for Boiling Water Reactors (BWR). The main idea is to obtain a Control Rod Pattern under the following considerations: (a) the critical reactor core state is satisfied, (b) the axial power distribution must be adjusted to a target axial power distribution proposal, and (c) the maximum Fraction of Critical Power Ratio (MFLCPR), the maximum Fraction of Linear Power Density (FLPD) and the maximum Fraction of Average Planar Power Density (MPGR) must be fulfilled. Those parameters were obtained using the 3D CM-PRESTO code. In order to decrease the problem complexity, Control Cell Core load strategy was implemented; in the same way, intermediate axial positions and core eighth symmetry were took into account. In this work, the cycle length was divided in 12 burnup steps. The Fuel Loading Pattern is an input data and it remains without changes during the iterative process. The Huitzoctli system was developed to use the combinatorial heuristics techniques Scatter Search and Tabu Search. The first one was used as a global search method and the second one as a local search method. The Control Rod Patterns obtained with the Huitzoctli system were compared to other Control Rod Patterns designs obtained with other optimization techniques, under the same operating conditions. The results show a good performance of the system. In all cases the thermal limits were satisfied, and the axial power distribution was adjusted to the target axial power distribution almost

  10. BWR type nuclear power plant and operation method therefor and method of forming oxide membrane on the surface of the constitutional member in contact with water

    International Nuclear Information System (INIS)

    In a BWR type nuclear power plant, an oxide membrane is formed on the surface of the constitutional members of a reactor primary system to be in contact with water while keeping the reactor water at a pH of 7.5 or less based on a room temperature and keeping a temperature of reactor water at 250degC or higher for 250 hours or more and then adding alkaline water to control the pH within a range of from 7.5 to 9.0 based on the room temperature and keeping the reactor water temperature to 250degC or higher for 100 hours or more. This process is conducted during the reactor shut down state and during the operation period from the time of the reactor shut down state to the time of the rated power operation state of the electric power generator. Then, a corrosion resistant oxide membrane with less involvement of radioactive ions can be formed, thereby enabling to improve corrosion resistance of nuclear fuel elements and suppressing the dose rate on the surface of pipelines of a primary coolant system, accordingly, operator's radiation dose rate can be reduced upon periodical inspection. (N.H.)

  11. Characterization of noble metals deposits and oxides in conditions of BWR reactors; Caracterizacion de depositos de metales nobles y oxidos en condiciones de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C.R.; Aguilar T, J.A.; Contreras R, A. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2008-07-01

    The oxides deposited on steel 304l under normal chemistry conditions (NWC) and hydrogen chemistry (HWC) with presence of Zn, being that the first ones present hexagonal oxides of Hematite and the second bipyramidal crystals possibly Magnetite with traces of Zn. Deposits of Pt on the oxidized surfaces under NWC conditions were obtained, being glasses from 2 to 4 {mu}m and Pt-Rh deposits were obtained on the oxidized surfaces in presence of Zn, by its size its were not possible to observe them by scanning electron microscopy. The kinetics of the surfaces of Hematite and of the deposits of Pt it was measured by means of the Tafel extrapolation technique, being proven the catalytic effect of the Pt, even in sizes of 11 {mu}m. (Author)

  12. 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)

  13. Boiling Water Reactor Loading Pattern Optimization Using Simple Linear Perturbation and Modified Tabu Search Methods

    International Nuclear Information System (INIS)

    An automated system for designing a loading pattern (LP) for boiling water reactors (BWRs) given a reference LP and control rod (CR) sequence has been developed. This system employs the advanced nodal code SIMULATE-3 and a BWR LP optimization code FINELOAD-3, which uses a simple linear perturbation method and a modified Tabu search method to select potential optimized LP candidates. Both of these unique methods of FINELOAD-3 were developed to achieve an effective BWR LP optimization strategy and to have high computational efficiency. FINELOAD-3 also adjusts deep CR positions to compensate for the core reactivity deviation caused by fuel shuffling. The objective function is to maximize the end-of-cycle core reactivity while satisfying the specified thermal margins and cold shutdown margin constraints. This optimization system realized the practical application for real BWR LP design. Computer time needed to obtain an optimized LP for a typical BWR/5 octant core with 15 depletion steps is ∼4 h using an engineering workstation. This system was extensively tested for real BWR reload core designs and showed that the developed LPs using this system are equivalent or better than the manually optimized LPs

  14. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 {sup o}C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV

  15. General features of direct-cycle, supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR

  16. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network; Prediccion del factor local de potencia en celdas de combustible BWR mediante una red neuronal multicapas

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 La Marquesa, Ocoyoacac, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2007-07-01

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U{sup 235}, some of these bars also contain a concentration of Gd{sub 2}O{sub 3} and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  17. Investigation of incident scenarios for BWR-type plant, taking into consideration the effects of the plant emergency protection systems. The Kruemmel reactor station (KKK) as an example. [accident management]. Untersuchungen zu Ereignisablaeufen fuer SWR-Anlagen unter Einbeziehung von Massnahmen des anlageninternen Notfallschutzes am Beispiel des Kernkraftwerkes Kruemmel (KKK)

    Energy Technology Data Exchange (ETDEWEB)

    Erven, U.; Nolte, W.; Linden, J. v.; Weidlich, H.

    1988-01-01

    The hypothetical scenarios discussed are the following: (1) Failure of the main heat sink with simultaneous breakdown of after-heat removal from the condensation chamber. (2) Failure of the entire, not battery-supported a.c. supply system (station blackout). (3) Loss of coolant in the containment, with failure of the sump pumping system. The study in addition to the initial failure assumes failure of further systems (condensation chamber cooling, non-battery-supported standby power supply, containment sump pumping system). In all three cases, further emergency provisions are possible in order to prevent dry-out of the reactor core, or undue pressure buildup in the containment. These provisions are easily feasible, as analyses have shown that there is sufficient time for carrying out the planned emergency protection measures or the plant protection measures, namely: 27 hours in case of 'failure of main heat sink', at least 5 hours in case of 'station blackout', and at least 4 days in case of 'loss of coolant in the containment'. The frequencies of occurrence of inadmissible plant conditions are low that there is consensus on the national and international level that such cases are not relevant for risk assessment, and therefore cannot be taken into consideration for technical protection measures development.

  18. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  19. Reactor physics calculations in the Nordic countries

    International Nuclear Information System (INIS)

    The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented

  20. Actinides record, power calculations and activity for present isotopes in the spent fuel of a BWR; Historial de actinidos y calculos de potencia y actividad para isotopos presentes en el combustible gastado de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Lucatero, M. A., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The administration of spent fuel is one of the more important stages of the nuclear fuel cycle, and this has become a problem of supreme importance in countries that possess nuclear reactors. Due to this in this work, the study on the actinides record and present fission products to the discharge of the irradiated fuel in a light water reactor type BWR is shown, to quantify the power and activity that emit to the discharge and during the cooling time. The analysis was realized on a fuel assembly type 10 x 10 with an enrichment average of 3.69 wt % in U-235 and the assembly simulation assumes four cycles of operation of 18 months each one and presents an exposition of 47 G Wd/Tm to the discharge. The module OrigenArp of the Scale 6 code is the computation tool used for the assembly simulation and to obtain the results on the actinides record presents to the fuel discharge. The study covers the following points: a) Obtaining of the plutonium vector used in the fuel production of mixed oxides, and b) Power calculation and activity for present actinides to the discharge. The results presented in this work, correspond at the same time immediate of discharge (0 years) and to a cooling stage in the irradiated fuel pool (5 years). (Author)

  1. A synergistic use of CFD, experiments and effective convectivity model to reduce uncertainty in BWR severe accident analysis

    International Nuclear Information System (INIS)

    In a previous work we presented an analysis approach developed to effectively and accurately assess thermal loads on vessel and structures in a Boiling Water Reactor (BWR) lower head during a severe accident. Central to the assessment is the Effective Convectivity Model (ECM) that makes use of experimental heat transfer correlations to capture the effect of turbulent natural convection in a volumetrically heated liquid pool, while retaining the pool three-dimensional energy splitting and ability to represent local heat transfer effects. Thanking to its features, the ECM is unique in enabling calculations of complex heat transfer phenomena during long severe accident transients that would not be otherwise feasible using higher-fidelity methods such as Computational Fluid Dynamics (CFD). Efficiency notwithstanding, the natural questions are: (i) how good are those ECM-calculated results, and, (ii) if required, what can be done (with the highest return-on-investment) to improve the quality of ECM prediction results. The approach refers to experiments and CFD simulations as the main resources to address (i) and (ii). However, validation of ECM against simulant-fluid experiments by itself does not reveal deficiencies (due to non-prototypicality factors). In the present work we focus on the use of CFD-based numerical 'experiments' to identify and quantify source of epistemic uncertainty in the calculated thermal loads due to modeling assumptions in ECM. Specifically, heat transfer correlations that underlie the ECM are obtained as surface-averaged (even though implemented as spatially distributed) and derived from experiments conducted at different geometries and using fluids that are not reactor prototypical (molten corium in the present case of severe accident). The CFD simulations exhibit so-called fluid Prandtl number effect on local peaking of the pool's downward heat flux for corium as working fluid. The main premise is a synergistic use of a fast-running model

  2. Panorámica de la actual narrativa colombiana: novedades y tendencias

    Directory of Open Access Journals (Sweden)

    Jorge Iván Parra

    2015-11-01

    Full Text Available Este artículo es fundamentalmente una visión panorámica de lo que en los últimos treinta años, ha producido la narrativa (por no decir, la novelística colombiana. La propuesta conduce la mirada hacia cuatro tendencias: histórica, urbana, de la violencia e intimista. Se entiende que una obra o un autor, eventualmente pueden, según sus características, acomodarse en varias tendencias a la vez, o que un autor puede ser visto en unas obras en una tendencia y en algunas obras en otra. Dado que este texto obedece a una apuesta personal producto de la relación directa lector-obras, prácticamente sin ninguna mediación, se inscribe dentro de lo que Edward Said cataloga como crítica práctica y de ahí el tenor de reseña en muchos de los comentarios.

  3. Organización y tendencias del conocimiento de enfermería en Chile

    Directory of Open Access Journals (Sweden)

    Sara Mendoza Parra

    2004-04-01

    Full Text Available Estudio cuantitativo, descriptivo y retrospectivo que exploró la organización del conocimiento de la Enfermería Chilena y sus tendencias. El universo estuvo constituido por reportes científicos publicados en las revistas de enfermería chilenas entre 1965 y 2003 (N=214. Se utilizó pauta basada en la CIPE, criterios de Nogueira e indicaciones de cienciometría. El análisis con medidas estadísticas de tendencia central fueron manejadas con SPSS. Algunos resultados fueron: el sujeto de estudio más frecuente fue el profesional de enfermería y la tendencia es enfocarse a las necesidades en salud de las personas adultas y en riesgo de enfermar. Se observa: las publicaciones tienen poco sustento teórico de enfermería, "Ciencia y Enfermería" es la revista que exhibe la mejor calidad científica.

  4. 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

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. Development of an Input Model to MELCOR 1.8.5 for the Oskarshamn 3 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Lars [Lentek, Nykoeping (Sweden)

    2006-05-15

    An input model has been prepared to the code MELCOR 1.8.5 for the Swedish Oskarshamn 3 Boiling Water Reactor (O3). This report describes the modelling work and the various files which comprise the input deck. Input data are mainly based on original drawings and system descriptions made available by courtesy of OKG AB. Comparison and check of some primary system data were made against an O3 input file to the SCDAP/RELAP5 code that was used in the SARA project. Useful information was also obtained from the FSAR (Final Safety Analysis Report) for O3 and the SKI report '2003 Stoerningshandboken BWR'. The input models the O3 reactor at its current state with the operating power of 3300 MW{sub th}. One aim with this work is that the MELCOR input could also be used for power upgrading studies. All fuel assemblies are thus assumed to consist of the new Westinghouse-Atom's SVEA-96 Optima2 fuel. MELCOR is a severe accident code developed by Sandia National Laboratory under contract from the U.S. Nuclear Regulatory Commission (NRC). MELCOR is a successor to STCP (Source Term Code Package) and has thus a long evolutionary history. The input described here is adapted to the latest version 1.8.5 available when the work began. It was released the year 2000, but a new version 1.8.6 was distributed recently. Conversion to the new version is recommended. (During the writing of this report still another code version, MELCOR 2.0, has been announced to be released within short.) In version 1.8.5 there is an option to describe the accident progression in the lower plenum and the melt-through of the reactor vessel bottom in more detail by use of the Bottom Head (BH) package developed by Oak Ridge National Laboratory especially for BWRs. This is in addition to the ordinary MELCOR COR package. Since problems arose running with the BH input two versions of the O3 input deck were produced, a NONBH and a BH deck. The BH package is no longer a separate package in the new 1

  10. 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

  11. 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.

  12. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.)

  13. Reactor water level control system

    International Nuclear Information System (INIS)

    A BWR type reactor comprises a control valve disposed in a reactor water draining pipelines and undergoing an instruction to control the opening degree, an operation board having a setting device for generating the instruction and a control board for giving the instruction generated by the setting device to the control valve. The instruction is supplied from the setting device to the control valve by way of a control circuit to adjust the opening degree of the control valve thereby controlling the water level in the reactor. In addition, a controller generating an instruction independent of the setting device and a signal transmission channel for signal-transmitting the instruction independent of the control circuit are disposed, to connect the controller electrically to the signal transmission. The signal transmission channel and the control circuit are electrically connected to the control valve switchably with each other. Since instruction can be given to the control valve even at a periodical inspection or modification when the setting device and the control circuit can not be used, the reactor water level can be controlled automatically. Then, operator's working efficiency upon inspection can be improved remarkably. (N.H.)

  14. Annual progress report for 1982 of Theoretical Reactor Physics Section

    International Nuclear Information System (INIS)

    The progress of work done in the Theoretical Reactor Physics Section of the Bhabha Atomic Research Centre, Bombay, during the calendar year 1982 is reported in the form of write-ups and summaries. The main thrust of the work has been to master the neutronic design technology of four different types of nuclear reactor types, namely, pressurized heavy water reactors, boiling light water reactors, pressurized light water reactors and fast breeder reactors. The development work for the neutronic analysis, fuel design, and fuel management of the BWR type reactors of the Tarapur Atomic Power Station has been completed. A new reactor simulator system for PHWR design analysis and core follow-up was completed. Three dimensional static analysis codes based on nodal and finite element methods for the design work of larger size (500-750 MWe) reactors have been developed. Space link kinetics codes in one, two and three dimensions for above-mentioned reactor systems have been written and validated. Fast reactor core disruptive analysis codes have been developed. In the course of R and D work concerning various types of reactor projects, investigations were also carried in the allied areas of Monte Carlo techniques, integral transform methods, path integral methods, high spin states in heavy nuclei and a hydrodynamics model for a laser driven fusion system. (M.G.B.)

  15. Overview of New Tools to Perform Safety Analysis: BWR Station Black Out Test Case

    Energy Technology Data Exchange (ETDEWEB)

    D. Mandelli; C. Smith; T. Riley; J. Nielsen; J. Schroeder; C. Rabiti; A. Alfonsi; Cogliati; R. Kinoshita; V. Pasucci; B. Wang; D. Maljovec

    2014-06-01

    Dynamic Probabilistic Risk Assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). While system simulator codes accurately model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by: 1) sampling values of a set of parameters from the uncertainty space of interest (using the simulation controller codes), and 2) simulating the system behavior for that specific set of parameter values (using the system simulator codes). For complex systems, one of the major challenges in using DPRA methodologies is to analyze the large amount of information (i.e., large number of scenarios ) generated, where clustering techniques are typically employed to allow users to better organize and interpret the data. In this paper, we focus on the analysis of a nuclear simulation dataset that is part of the Risk Informed Safety Margin Characterization (RISMC) Boiling Water Reactor (BWR) station blackout (SBO) case study. We apply a software tool that provides the domain experts with an interactive analysis and visualization environment for understanding the structures of such high-dimensional nuclear simulation datasets. Our tool encodes traditional and topology-based clustering techniques, where the latter partitions the data points into clusters based on their uniform gradient flow behavior. We demonstrate through our case study that both types of clustering techniques complement each other in bringing enhanced structural understanding of the data.

  16. Numerical analysis of the mixing and recombination in the downcomer of an internal pump BWR

    International Nuclear Information System (INIS)

    The mixing process of feedwater and reactor water in the downcomer of an internal-pump BWR (Forsmark 1 and 2) has been numerically modelled by means of a CFD-code (FLUENT/UNS). Earlier studies with a very rough model, have shown that a new sparger design is necessary to achieve an effective HWC through improved mixing in the downcomer,. This requires detailed and accurate modelling of the flow, not only for determining the mixing quality but for avoiding negative effects like increased thermal loading of internal parts. Through three 22.5deg models containing a sparger end and half the region between spargers, the principles of a new design have been defined. Their length scales range from 7-14 mm to ca 12 m. Also the steam separator region has been incorporated in the models. A 90deg model shows that they are sufficiently accurate for the actual region. The results cannot be generalised to other regions between spargers due to geometrical differences affecting the flow and the mixing. The spargers have to be lengthened to ensure a complete coverage of all the downcomer. This condition is necessary but not sufficient since the lengthening is accompanied by an unfavourable modification of the flow. However, a reduction of the sparger vertical size and a front with fewer but larger holes extensively improves the mixing. To confirm that improvement in mixing is accompanied by one in recombination conditions, a number of fluid particles have been tracked in the downcomer. The results correlate well with the temperature field at the lower plane. (author)

  17. Design of a fuel recharge for a BWR using advanced optimization systems

    International Nuclear Information System (INIS)

    The fuel recharge design for a BWR reactor it was carried out, which includes the design of four fuel cells to form an assembly, the accommodation design of fresh and partially consumed assemblies and the control bars pattern design to use along an operation cycle. The three stages were approached as optimization problems using different computational tools, each one of those includes an objective function to measure quantitatively the evolution of the different candidate solutions. With the tool used in the fuel cells design that makes use of the tabu search technique its were obtained cells that showed to be lightly more reactive that other similar taken as reference. With the four designed cells it was formed a fuel assembly that turn out to have an average enrichment lightly smaller to the one of another assembly similar taken as reference. In the recharge pattern design it was used another optimization tool, also based on tabu search to obtain the accommodation of 108 fresh fuels and 336 partially consumed, fulfilling the conditions imposed to operate with the core strategies with control cells (CCC) and of low leakage. A Haling calculation reported that with the obtained accommodation it was achievement to increase in 8% the cycle length with regard to the one obtained using a similar reference pattern. In the design of the control bars patterns it was used a tool based on the use of the genetic algorithms to obtain the placement patterns of the control bars along an operation cycle. The search tool only uses the bars of the A2 sequence and it makes use of the 1/8 symmetry of the core, with that the number of used control bars it decreases at 5. Also the use of control bars in intermediate positions is also avoided. With the obtained patterns a cycle length is obtained that is lightly bigger to the reported value in a Haling calculation. (Author)

  18. Experimental investigations of BWR pressure suppression pool behavior under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    The experiments discussed in this paper look into different processes which may occur during a loss-of-coolant accident in the pressure suppression pool of a Boiling Water Reactor (BWR). These processes include: a) development of a thermal stratification, b) bubble dynamics and related water flow during continuous release of air and c) air blowdown and associated water slug phenomenon in the water pool. The experiments have been performed in the THAI test facility, which is a cylindrical vessel of 9.2 m height, 3.2 m diameter and with a gas volume of 60 m3. The variation in the investigated test parameters included, steam and air mass flux, initial water pool temperature, blowdown pressures, downcomer submergence, etc. A systematic variation of the test parameters allowed better understanding of the phenomena. Experiments discussed in this paper were performed with a vertical downcomer of 0.1 m diameter and 2 m submergence depth in the water pool. For the blowdown experiments, a separate interconnecting vessel of 1 m3 volume was used to inject air at pressures between 3 bar and 10 bar. A high speed camera (1000 fps) was installed to visualize the formation and propagation of air bubbles in the suppression pool and the resulting pool swelling phenomena. Customized instrumentation applied during the tests included grids of densely spaced thermocouples and of pressure transducers at various locations in order to capture the temperature distribution in the pool and the water slug induced pressure loadings, respectively. The present paper discusses the main outcome of the selected experiments. On the whole the experimental data may be very useful for code validation. (authors)

  19. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

    International Nuclear Information System (INIS)

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

  20. TRAC-BF1/NEM stability methodology for BWR core wide and regional stability analysis

    International Nuclear Information System (INIS)

    A time-series analysis stability methodology is presented based on the TRAC-BF1/NEM coupled code. The methodology presented has a potential application for BWR core-wide and regional stability studies allowed by the 3D capabilities of the code. The stability analysis is performed at two different levels: using the TRAC-BF1 point kinetics model and employing the three-dimensional neutronic transient capability of the NEM code. Point kinetics calculations show power fluctuations when white noise is applied to the inlet mass flow rate of each of the channel components. These fluctuations contain information about the system stability, and are subsequently studied with time-series analysis methods. The analysis performed showed that the reactor core has a low-frequency resonance typical of BWRs. Analysis of preliminary three-dimensional calculations indicates that the power fluctuations do not contain the typical resonance at low frequency. This fact may be related to the limitation of the thermal-hydraulic (T-H) feedback representation through the use of two-dimensional tables for the cross-sections needed for 3D kinetics calculations. The results suggest that a more accurate table look-up should be used, which includes a three-dimensional representation of the feedback parameters (namely, average fuel temperature, average moderator temperature, and void fraction of the T-H cell of interest). Further research is being conducted on improving the cross-section modeling methodology, used to feed the neutron kinetics code for both steady state and transient cases. Also a comprehensive analysis of the code transient solution is being conducted to investigate the nature of the weak dependence of the power response on T-H variations during the performed 3D stability transient calculations

  1. Evaluation of silica behavior for reducing the precoating frequency of the reactor water cleanup system of the Hamaoka NPS

    International Nuclear Information System (INIS)

    When the silica concentration in reactor water exceeds the reference value, the ion-exchange resin powder used in the reactor water cleanup system (CUW) is replaced with new resin powder. This is referred to as the 'precoating of CUW'. Precoating of CUW generates radioactive waste; therefore, a higher frequency of CUW precoating increases the waste disposal cost. In the Advanced Boiling-Water Reactor (ABWR) of Hamaoka Unit 5, the frequency of CUW precoating has been higher than that of any other plants as a result of the high silica concentration in the reactor water; therefore, the behavior of silica in the reactor primary water circuit was examined in order to reduce the frequency of CUW precoating. A calculation model was developed for the silica behavior in the primary water (Silica Behavior Code) and the mass balances of silica in Hamaoka Units 3 and 4 (BWR5), and Unit 5 was then analyzed applying this code. A comparison of these mass balances shows two results for the difference between BWR5 and the ABWR. First, the amount of silica removed from the condensate demineralizer (CD) in the ABWR is less than that in BWR5, because silica, which is transported from the reactor water into the main steam, bypasses the CD and returns to the reactor water directly due to the high pressure heater drain line specific to the ABWR. Second, the amount of silica generated during plant operation in the ABWR is greater than that in BWR5 due to the generation of silica in the high pressure heater drain line. From the above results, it is concluded that the high pressure heater drain line, which is specific to the ABWR, is the cause of the high silica concentration in reactor water in Hamaoka Unit 5. (author)

  2. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  3. Performance of iron-chromium-aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    Science.gov (United States)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun; Heuser, Brent J.; Mandapaka, Kiran K.; Was, Gary S.

    2016-03-01

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

  4. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    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 U{sup 235} 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

  5. Needs of nuclear data for advanced light water reactor

    International Nuclear Information System (INIS)

    Hitachi has been developing medium sized ABWRs as a power source that features flexibility to meet various market needs, such as minimizing capital risks, providing a timely return on capital investments, etc. Basic design concepts of the medium sized ABWRs are 1) using the current ABWR design which has accumulated favorable construction and operation histories as a starting point; 2) utilizing standard BWR fuels which have been fabricated by proven technology; 3) achieving a rationalized design by suitably utilizing key components developed for large sized reactors. Development of the medium sized ABWRs has proceeded in a systematic, stepwise manner. The first step was to design an output scale for the 600MWe class reactor (ABWR-600), and the next step was to develop an uprating concept to extend this output scale to the 900MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. In addition, Hitachi and MHI developed an ultra small reactor, 'Package-Reactor'. About the nuclear data, for the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores. (author)

  6. 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

  7. Analysis of the behavior of irradiated BWR fuel rod in storage dry conditions; Analisis del comportamiento de una barra combustible irradiada BWR en condiciones de almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Montes, D.; Ruiz-Hervias, J.; Munoz-Reja, C.

    2014-07-01

    In order to complete previous studies of creep on PWR sheath material, developed a joint experimental program by CSN, ENRESA and ENUSA about BWR (Zircaloy-2) sheath material. This program consisted in creep tests and then on the material under creep, compression testing diametral obtaining the permissible displacement of the sheath to break. (Author)

  8. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional {delta}-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the {epsilon}/d model, was developed. The correct solution has been derived in the {epsilon}/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In

  9. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    International Nuclear Information System (INIS)

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  10. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  11. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [Argonne National Lab., IL (United States)] [and others

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  12. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

    International Nuclear Information System (INIS)

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  13. Comparison of heat capacity and thermal time constant between BWR fuel and simulated heater rod

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    It is important to know the thermal characteristics of BWR fuel, i.e. heat capacity and thermal time constant, in order to evaluate the thermal hydraulics at BWR accidents and the events under thermal-hydraulic and neutronic coupling condition. Further, since the heater rod simulating BWR fuel is used in the tests for BWR accidents and for BWR thermal hydraulics coupled with neutronics, it is important to know the thermal characteristics of the heater rod. Therefore, the author investigated the thermal characteristics of BWR fuel and the heater rod by performing experiments and analyzing with J-TRAC code capable to analyze 2-dimensional heat conduction problem. The heat capacity per unit length of BWR fuel cp{rho}A (kJ/mK) was estimated to be 0.34 kJ/mK - 0.36 kJ/mK in 300 deg. C - 800 deg. C. The heat capacity of the heater rod was almost identical with each other regardless of the differences in rods and positions. It was higher with higher temperature. The heat capacity of the heater rod used in the test for BWR accidents was about 0.38 kJ/mK at 600 deg. C, which was about 9% higher than the average (0.35 kJ/mK) of BWR fuel. On the other hand, the heat capacity used in the test for BWR thermal hydraulics coupled with neutronics was about 0.42 kJ/mK at 600 deg. C, which was about 20% higher than the average of BWR fuel. Thermal time constant was affected by surface heat transfer coefficient, thermal diffusivity, and gap conductance. When the surface heat transfer coefficient is small, it controls the heat transfer and thermal time constant depends mainly on the surface heat transfer coefficient. When the surface heat transfer coefficient is large, the heat conduction controls the heat transfer and thermal time constant depends mainly on the thermal diffusivity. In the former case, one point heat transfer model is applicable and the thermal time constant is proportional to the inverse of the surface heat transfer coefficient. In this case, the thermal time

  14. Evaluation of soil-structure interaction based on forced vibration tests of three reactor buildings within a site

    International Nuclear Information System (INIS)

    Three BWR type reactor buildings (Unit No. 1, 2, 3) have been constructed at Hamaoka site in Shizuoka prefecture, Japan, during the past fifteen years. In order to confirm and verify the seismic design procedure, forced vibration tests were performed for the completed three reactor buildings. Enough test data have been accumulated and consequently, it is now possible to consolidate these data and compare the differences of vibrational characteristics of the three reactor of soil-structure interaction effects through the simulation study of vibrational characteristics of the three reactor buildings and explores realistic vibration models

  15. Last experiences on ID BWR shroud inspection and the new developments to examine the below core plate areas

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.; Willke, A.; Yague, L. [TECNATOM SA, Madrid (Spain)

    2001-07-01

    In recent years, the owners of BWR type nuclear power plants have had to address new inspection requirements relating to the core shroud inside the reactor vessel, the aim of which is to contain the fuel assemblies and provide support for the structures located in the upper part of the reactor. The shroud consists of a cylinder measuring some 40-50 mm in thickness, manufactured from various sections of AISI-304 stainless steel and INCONEL, joined by vertical and circumferential welds. The appearance of unstable cracks in these welds would directly affect the structural integrity of the component and the safety of the plant. As regards access to the core shroud and to the surface to be examined, two alternatives might be considered: inspection from outside the component, moving along the so-called annulus between the reactor vessel wall and the component (OD inspection), or from the interior (ID inspection). With a view to addressing this problem, Tecnatom has in recent years launched several projects, grouped under the generic name TEIDE, in order to develop scanners and NDT techniques achieving the maximum inspection coverage of this component. The decision was taken to perform ID inspections, mainly because this type of scanners were not available at that time, and which provide the 4 following advantages. 1) Maximum inspected weld length. This avoids interference with the jet pumps and the systems present in the annulus and affecting OD inspections. Besides, the repairs performed on in-service core shrouds in all cases imply the addition of new fixed elements on their outer surface, since the fuel assembly space must be left free. 2) Reduction of inspection times and of unforeseen events: maintenance of planning schedules, reduction of personnel doses, reduced critical path time. 3) High inspection accuracy and repeatability. 4) Simplification of equipment positioning work (similar to the installation of fuel assemblies). As regards inspection techniques, the

  16. Tendencia de I+D+i en Biomecánica aplicada al deporte

    OpenAIRE

    Ferro Sanchez, Amelia

    2012-01-01

    Análisis de la investigación en el ámbito de la Biomecánica aplicada al deporte. La tendencia de los proyectos en I+D+i de la Biomecánica del Deporte ha cambiado notablemente en las dos últimas décadas. Los investigadores han pasado de aplicar la biomecánica al alto rendimiento deportivo a tratar temas transversales, relacionados en muchos casos con la salud, fisiología, rehabilitación, el control motor, la gestión y, entre otros, el diseño de equipamiento e instalaciones.

  17. Tendencias tecnológicas en Europa. Análisis de los procesos de prospectiva

    OpenAIRE

    Martín Pereda, José Antonio

    2003-01-01

    El Informe «Tendencias tecnológicas en Europa. Análisis de los procesos de prospectiva», que Cotec publica en su Colección Observaciones de Buenas Prácticas en los sistemas de innovación, es el resultado de un trabajo original e innovador que compara los resultados obtenidos en los estudios de prospectiva tecnológica realizados por cinco de los países más avanzados de Europa.

  18. Análisis de tendencias de heliofanía efectiva en Argentina

    OpenAIRE

    Hugo Grossi Gallegos; María Isabel Spreafichi

    2007-01-01

    Se presentan resultados del análisis estadístico de series temporales de promedios anuales de heliofanía efectiva (horas de brillo solar) registrados en estaciones de Argentina, que compensan la insuficiencia de datos de irradiación solar y permiten estudiar tendencias relacionadas con el Cambio Climático Global. Se aplicó un estudio previo de consistencia teniendo en cuenta el error de medición, con la aplicación del test de Mann-Kendall y se discuten los resultados obtenidos. Aparecen algun...

  19. La tendencia post gerencial en la determinación de la cultura organizacional del futuro

    OpenAIRE

    Lucas del Moral

    2000-01-01

    Este artículo nace de la necesidad de superar la falta de definición conceptual de lo que enfrentaremos en el futuro casi inmediato en el mundo de la administración de empresas y tuvo como finalidad la revisión bibliográfica desde el punto de vista del método de investigación documental-cualitativa, sobre una de las recientes tendencias de la Gerencia. Enfoca la atención sobre las tareas pasadas, actuales y futuras del Management y su influencia contemporánea en la determinación de la cultura...

  20. Tendencia temporal de la mortalidad por armas de fuego en Argentina, 1980-2012

    OpenAIRE

    Hugo Spinelli; Adrián Santoro; Carlos Guevel; Marcio Alazraqui

    2015-01-01

    Este trabajo analiza el impacto de las armas de fuego en la mortalidad entre los años 1980-2012 en Argentina. Para ello se realiza un estudio epidemiológico descriptivo de la tendencia temporal, a través de las siguientes variables: sexo, grupos de edad, intencionalidad y jurisdicción. Los datos fueron obtenidos en la Dirección de Estadísticas e Información en Salud del Ministerio de Salud de la Nación. Se encontraron 87.671 muertes por armas de fuego. Los varones representaron el 85,7% y su ...

  1. Tendencias de mortalidad por traumatismos y envenenamientos en adolescentes: México, 1979-1997

    OpenAIRE

    Celis Alfredo; Gómez-Lomelí Zoila; Armas Jesús

    2003-01-01

    OBJETIVO: Describir las causas externas de la mortalidad por traumatismos y envenenamientos, así como la tendencia que estas causas han mostrado entre adolescentes en México, de 1979 a 1997. MATERIAL Y MÉTODOS: Para la realización de este estudio descriptivo de la mortalidad se utilizaron las bases de datos de mortalidad registrada en México de 1979 a 1997. Los datos son recabados, codificados y capturados por el Instituto Nacional de Estadística, Geografía e Informática, y se obtuvieron a tr...

  2. Algunas tendencias jurídicas de la globalización

    OpenAIRE

    De Miguel Asensio, Pedro Alberto

    2003-01-01

    La globalización se halla asociada a nuevas tendencias jurídicas en la ordenación de las actividades de sus actores, muy especialmente las empresas transnacionales, pero también el resto de las empresas y los particulares, así como de las relaciones de esos actores con los Estados. Objeto de especial atención son: el progresivo desarrollo de nuevos ámbitos de elaboración de normas, la creciente competencia de localización entre ordenamientos y el papel que desempeñan las organizadas privadas ...

  3. El suicidio y su tendencia social en México: 1990-2011

    OpenAIRE

    René Alejandro JIMÉNEZ-ORNELAS; Leticia CARDIEL-TÉLLEZ

    2013-01-01

    En las últimas dos décadas, al igual que en el resto del mundo, las tasas de mortalidad por sui - cidio en México han presentado una constante tendencia al incremento. Esto ha catalogado al suicidio como un problema de salud que debe ser analizado desde un enfoque multidimensional, para así examinar las características socio demográficas de los suicidas que permitan generar medidas preventivas ante dicha situación y, por ende, procurar la disminución de la incidencia de este fenómeno que se e...

  4. Estado y tendencia de la Administración de Proyectos en México

    OpenAIRE

    Vega Díaz, José Alberto

    2004-01-01

    El objetivo general de este estudio, es analizar el estado y las tendencias de la Administración de Proyectos en México e identificar el nivel de harmonización que hay entre la teoría y la puesta en práctica dentro de la industria mexicana. Esto ayudará a las partes interesadas a evaluar el estado del uso de la Administración de Proyectos dentro de su propia compañía en una base comparativa, reconocer y discutir las necesidades que tiene la práctica de la Administración d...

  5. Reduction of the consequences of accidents whereby the emergency shutdown system in modern reactors fails (ATWS)

    International Nuclear Information System (INIS)

    If a nuclear reactor can not be shutdown by pulling out the control rods, an emergency shutdown system must be used. The events, when such a system fails, have been calculated. Also attention is paid to the chance that both systems fail and the possibility of using an extra independent shutdown system, realized in pressurized water reactors (PWR) or boiling water reactors (BWR). Finally a General Electric developed safety method and an alternative method regarding the failure of an emergency shutdown system are described. The results of this investigation, which were also based on a literature study, can be applied in formulating specifications of new nuclear power plants

  6. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  7. Safety Evalution Report related to the final design approval of the GESSAR II BWR/6 nuclear island design (Docket No. 50-447)

    International Nuclear Information System (INIS)

    This report supplements the GESSAR II SER (NUREG-0979), issued in April 1983, summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design. The review is carried out in accordance with the procedures for demonstrating the acceptability of the design for the severe-accident concerns described in draft NUREG-1070, NRC Policy on Future Reactor Designs: Decisions on Severe Accident Issues in Nuclear Power Plant Regulation. Supplement 2 also provides more recent information regarding resolution or update of the confirmatory items and FDA-1 conditions identified in SSER 1. Subject to favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in draft NUREG-1070

  8. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  9. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 13. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  10. 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)

  11. 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)

  12. 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)

  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 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)

  14. 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)

  15. 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

  16. BDDR, a new CEA technological and operating reactor database

    Energy Technology Data Exchange (ETDEWEB)

    Soldevilla, M.; Salmons, S.; Espinosa, B. [CEA-Saclay, CEA/DEN/DANS/DM2S/SERMA, 91191 Gif-sur-Yvette (France); Clanet, M.; Boudin, X. [CEA-Bruyeres-le-Chatel, 91297 Arpajon (France)

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  17. 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)

  18. Calibration of the enigma code for Finnish reactor fuel with support from experimental irradiations

    International Nuclear Information System (INIS)

    Assessment by VTT of the ENIGMA fuel performance code, the original version by Nuclear Electric plc of the UK amended by a set of WWER specific materials correlations, is described. The given examples of results include analyses for BWR 9 x 9 fuel, BWR fuel irradiated in the reinstrumented test of an international Riso project, pre-characterized commercial WWER fuel irradiated in Loviisa reactor in Finland, and instrumented WWER test fuel irradiations in the MR reactor in Russia. The effects of power uncertainty and some model parameters are discussed. Considering the fact that the described cases all mean prototypic application of the code, the results are well encouraging. The importance of the accuracy in temperature calculations is emphasized. (author). 2 refs, 12 figs, 1 tab

  19. Tendencias de la Educación universitaria en el siglo XXI

    Directory of Open Access Journals (Sweden)

    Ortega Castro, Vicente

    2002-09-01

    Full Text Available Not available

    El cambio de milenio ha propiciado muchos informes, libros y conferencias sobre la situación de las Universidades y los posibles cambios que se avecinaban como consecuencia de las tendencias de cambios en la sociedad, entre los que se encuentran el predominio del conocimiento, laglobalización y la influencia de nuevas tecnologías aplicadas al ámbito de la educación. Para la sociedad europea, en la que España está inmersa política, social y económicamente, es especialmente relevante la expectativa que está despertando la Declaración de Bolonia. Pues bien, tomando este documento como referencia principal, este artículo examina las funciones permanentes de la Universidad: creación de conocimientos, formación de profesionales, servicio a la sociedad y otros, teniendo en cuenta la importancia de la calidad, de la internacionalización, de la competitividad, de la movilidad, de la complejidad, de las nuevas demandas, de los nuevos esquemas de financiación y otros factores que hacen necesario que las Universidades comprendan las nuevas tendencia y hagan un esfuerzo, no solo de adaptación, sino de anticipación.

  20. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices