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Sample records for reactor pbmr basado

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  2. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  3. Nuclear reactor PBMR and cogeneration

    International Nuclear Information System (INIS)

    Ramirez S, J. R.; Alonso V, G.

    2013-10-01

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

  4. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  5. Optimisation of Deep Burn Incineration of Reactor Waste Plutonium in a PBMR DPP-400 core

    International Nuclear Information System (INIS)

    Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

    2013-01-01

    The incineration of pure Pu in the PBMR-400, using the simulated fuel sphere geometries, are not recommended, since it violated the safety limits. The addition of MA to Pu(PWR) is not recommended since this substantially increased the mass of heavy metals to be disposed in the spent fuel. Therefore a redesign of both the reactor and fuel sphere geometries are recommended, in order to resolve the safety issues

  6. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    Directory of Open Access Journals (Sweden)

    Ralind Re Marla

    2015-03-01

    Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

  7. Risk, probability and uncertainty in the calculations of gas cooled reactor of PBMR type. Part 2

    International Nuclear Information System (INIS)

    Serbanescu, Dan

    2004-01-01

    The paper presents the main conclusions of the insights to a cooled gas reactor from the perspective of the following notions: probability, uncertainty, entropy and risk. Some results of the on-going comparison between the insights obtained from three models and approaches are presented. The approaches consider the Pebble Bed Module Reactor (PBMR) NPP as a thermodynamic installation and as hierarchical system with or without considering the information exchange between its various levels. The existing model was a basis for a PRA going on in phases for PBMR. In the first part of this paper results from phase II of this PRA were presented. Further activities going on in the preparation for phase II PRA and for the development of a specific application of using PRA during the design phases for PBMR are undergoing with some preliminary results and conclusions. However, for the purposes of this paper and the comparative review of various models in the part two one presents the risk model (model B) based on the assumption and ideas laid down at the basis of the future inter-comparison of this model with other plant models. The assumptions concern: the uncertainties for the quantification of frequencies; list of initiated events; interfaces with the deterministic calculation; integrated evaluation of all the plant states; risk of the release of radionuclide; the balance between the number and function of the active systems and the passive systems; systems interdependencies in PBMR PRA; use of PRA for the evaluation of the impact of various design changes on plant risk. The model B allows basically evaluating the level of risk of the plant by calculating it as a result of acceptance challenge to the plant. By using this model the departure from a reference state is given by the variation in the risk metrics adopted for the study. The paper present also the synergetic model (model C). The evaluation of risk in the model C is considering also the information process. The

  8. A CFD method to evaluate the integrated influence of leakage and bypass flows on the PBMR Reactor Unit

    International Nuclear Information System (INIS)

    Janse van Rensburg, J.J.; Kleingeld, M.

    2010-01-01

    Research highlights: → Research and analysis to identify and rank different leakage flow paths in a HTR. → Development of integrated CFD methodology for the prediction of leakage flows. → Development of a methodology to simulate flow resistances in above CFD model. → Validation of predicted flow results against different numerical methodology. → Illustration of the significant improvement achieved through this methodology. - Abstract: An area that has been identified as significantly important in the development of a High Temperature Reactor (HTR) is the prediction of leakage and bypass flows through such a reactor. It is therefore essential to understand the causes of bypass flows and to determine the effect on the predicted fuel and component temperatures. This paper discusses the identification of leakage flows that are applicable to the Pebble Bed Modular Reactor (Pty) Ltd. (PBMR) design and the ranking of these leakage flows. The modeling methodology and results are also discussed. Similar to previous HTR's, it was found that leakage and bypass flows are important parameters to consider for safe and efficient operation of the PBMR. Through a focused approach, it is shown that PBMR is able to improve the understanding of this phenomenon and quantify the flows and subsequent influence on the operation of the system. This has resulted in a reduction of leakage and bypass from approximately 46% to 20%. The improved understanding of leakage and bypass flows allows PBMR to address this issue during the design phase of the project, which subsequently results in a vast improvement over historical HTR designs. This gives PBMR a distinct advantage over previous High Temperature Reactors.

  9. Conceptual design of a nucleo electric simulator with PBMR reactor based in Reduced order models

    International Nuclear Information System (INIS)

    Valle H, J.; Morales S, J.B.

    2005-01-01

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

  10. Optimisation of deep burn incineration of reactor waste plutonium in a PBMR DPP-400 core

    International Nuclear Information System (INIS)

    Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

    2014-01-01

    In this article an original set of coupled neutronics and thermo-hydraulic simulation results for the VSOP 99/05 diffusion code are presented for advanced fuel cycles for the incineration of weapons-grade plutonium, reactor-grade plutonium and reactor-grade plutonium with its associated Minor Actinides in the 400 MW th Pebble Bed Modular Reactor Demonstration Power Plant. These results are also compared to those of the standard 9.6 wt% enriched 9 g/fuel sphere U/Pu fuel cycle. The weapons-grade and reactor-grade plutonium fuel cycles produced good burn-ups. However, the addition of the Minor Actinides to the reactor-grade plutonium caused a large decrease in the burn-up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which was intended for direct disposal in a deep geological repository, without chemical reprocessing. All the plutonium fuel cycles failed the adopted safety limits used in the PBMR400 in that either the maximum fuel temperature of 1130 °C during normal operation, or the maximum power density of 4.5 kW/sphere was exceeded. All the plutonium fuel cycles also produced positive uniform temperature reactivity coefficients, i.e. the reactivity coefficient where the temperatures of the fuel and the graphite moderator in the fuel spheres were varied together. These unacceptable positive coefficients were experienced at low temperatures, typically below 700 °C. This was due to the influence of the thermal fission cross-section resonances of 239 Pu and 241 Pu. Weapons-grade plutonium produced the worst safety performance. The safety performance of the reactor-grade plutonium also deteriorated when the HM loading was reduced from 3 g/sphere to 2 g or 1 g

  11. Optimisation of deep burn incineration of reactor waste plutonium in a PBMR DPP-400 core

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School for Mechanical and Nuclear Engineering, North West University, PUK-Campus, Private Bag X6001, Internal Post Box 360, Potchefstroom 2520 (South Africa); Mulder, Eben J. [School for Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

    2014-05-01

    In this article an original set of coupled neutronics and thermo-hydraulic simulation results for the VSOP 99/05 diffusion code are presented for advanced fuel cycles for the incineration of weapons-grade plutonium, reactor-grade plutonium and reactor-grade plutonium with its associated Minor Actinides in the 400 MW{sub th} Pebble Bed Modular Reactor Demonstration Power Plant. These results are also compared to those of the standard 9.6 wt% enriched 9 g/fuel sphere U/Pu fuel cycle. The weapons-grade and reactor-grade plutonium fuel cycles produced good burn-ups. However, the addition of the Minor Actinides to the reactor-grade plutonium caused a large decrease in the burn-up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which was intended for direct disposal in a deep geological repository, without chemical reprocessing. All the plutonium fuel cycles failed the adopted safety limits used in the PBMR400 in that either the maximum fuel temperature of 1130 °C during normal operation, or the maximum power density of 4.5 kW/sphere was exceeded. All the plutonium fuel cycles also produced positive uniform temperature reactivity coefficients, i.e. the reactivity coefficient where the temperatures of the fuel and the graphite moderator in the fuel spheres were varied together. These unacceptable positive coefficients were experienced at low temperatures, typically below 700 °C. This was due to the influence of the thermal fission cross-section resonances of {sup 239}Pu and {sup 241}Pu. Weapons-grade plutonium produced the worst safety performance. The safety performance of the reactor-grade plutonium also deteriorated when the HM loading was reduced from 3 g/sphere to 2 g or 1 g.

  12. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR

    International Nuclear Information System (INIS)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L.

    2008-01-01

    In this work was developed the neutronic model of a fuel sphere of a nuclear reactor of gas of high temperature to modulate of bed of spheres (PBMR), using the Monte Carlo method with the MCNPx code. In order to be able to verify the fuel model constructed in this investigation, it is used a case of reference, based on an international exercise b enchmark . The benchmark report contains the results sent by different international participants for five phases with respect to the high temperature gas reactor (HTR), fed with uranium, plutonium and thorium. In particular, in first stage of benchmark an infinite adjustment of uranium compound fuel spheres is considered unique, with which our results were compared. This first stage considers two cases: cell calculations with spherical external frontier and cell calculations with cubic external frontier. The objective is to identify any increase in the uncertainty, related to the uranium fuel, that is associated with the plutonium and thorium fuels. In order to validate our results, the values of the neutron multiplication factor were taken in account, in cold and in the heat of the moment from the participants who sent their results obtained with Monte Carlo and deterministic calculations. The model of the fuel sphere developed in this work considers a regular distribution of 15000 Triso particles, in a cubic mesh centered within the sphere. For it was necessary to define the step firstly or p itch o f the cubic mesh. Generally, the results obtained by the participants of benchmark and those of this investigation present good agreement, nevertheless, appear some discrepancies, attributed to factors like different libraries of cross sections used, the nature of the solution: Monte Carlo or deterministic, and the difficulty of some participants to model the external frontier condition of reflection. (Author)

  13. The pebble bed modular reactor (PBMR) as a source of high quality process heat for sustainable oil sands expansion

    International Nuclear Information System (INIS)

    Morris, A.; Kuhr, R.

    2008-01-01

    Bitumen extraction, processing and upgrading consumes large quantities of natural gas for production of steam, hot water and hydrogen. Massive expansion of bitumen production is planned in response to energy demands, oil prices, and the desire for energy security. The PBMR in its Process Heat configuration supports applications that compete in a cost effective and environmentally sustainable way with natural gas fired boilers and steam methane reforming. The PBMR has the benefit of size, passive nuclear safety characteristics (encompassing Generation IV safety principles), high reliability, high temperature process heat (750-950 o C) in a modular design suited to the oil sands industry. (author)

  14. PBMR Project - Pebble Fuel Advantages

    International Nuclear Information System (INIS)

    Slabber, Johan; Matzie, Regis; Casperson, Sten; Kriel, Willem

    2006-01-01

    An overview is presented of all the important issues that influenced the choice of pebble fuel for the High-temperature Gas-cooled Reactor (HTGR) concept developed by South Africa. Each of these issues is then discussed in detail and compared with other fuel configurations proposed for direct cycle High-temperature Reactor (HTR) applications. The comparisons are provided using objective data generated by analyses done for the design of the Pebble Bed Modular Reactor (PBMR) and data that is available in open literature for the other fuel configurations

  15. Technical status of the pebble bed modular reactor (PBMR-SA) conceptual design

    International Nuclear Information System (INIS)

    Fox, M.

    1997-01-01

    The reactor study is well underway seen from a broad spectrum of disciplines and technology. The objective power output with a high efficiency direct cycle power conversion unit remains promising after compiling the first critical analysis of the core and the power conversion unit. The stability and controllability of the system are demonstrated by the engineering simulator. The main system and components are basically specified for costing purposes. A first plant layout has been completed demonstrating the positions of main components, personnel movement, installation methods for large components, etc. A cryptic report style presentation includes study objectives, indicating guiding documents, giving an overview of design and analyses work done as well as a few sketches and diagram are included in this paper. Most of these sketches and diagrams are small replicas of large drawings and are therefore not readable but can be used as references. (author)

  16. PBMR-SA licensing project organization

    International Nuclear Information System (INIS)

    Clapisson, G.A.; Metcalf, P.E.; Mysen, A.

    2001-01-01

    The South African nuclear regulatory authority, the Council for Nuclear Safety (CNS), is beginning the safety review of the Pebble Bed Modular Reactor (PBMR) design under development by the South African National Electrical Utility, Eskom. This paper describes the CNS licensing process, including the establishment of basic licensing criteria, general design criteria, and specific design rules, as well the safety assessment to be conducted in accordance with the established structure. It also summarises the CNS PBMR review project activities, including the overall organisational arrangements, licensing basis, safety and risk assessment, general operating rules and plant design engineering, and pre-operational testing. (author)

  17. Current status of the PBMR licensing project

    International Nuclear Information System (INIS)

    Mysen, A.; Clapisson, G.A.; Metcalf, P.E.

    2000-01-01

    The CNS is currently reviewing the PBMR conceptual design from a licensibility point of view. The PBMR concept is based on a High Temperature Gas Cooled Reactor - pebble bed reactor type. It is anticipated that the PBMR design will rely on inherent safety characteristics to contain fission products within fuel over the full range of design basis events. This feature combined with the high temperature integrity of the fuel and structural graphite, allows the safe use of a high coolant temperature, which allows consideration of the future development of this reactor for non-electrical applications of nuclear heat for industrial use. The CNS licensing approach requires that the licensing and design basis of the plant should respect prevailing international norms and practices and that a quantitative risk assessment should demonstrate compliance with the CNS fundamental safety standards. The first stage of the licensing process is now ongoing; this is a pre-application phase, which will result in a statement on licensibility being issued. Identification of the specific documentation requirements and information needed is required across every step of the licensing process. Top level regulatory requirements have been established for the PBMR. They include the CNS fundamental safety standard and basic licensing criteria, which describes requirements on licensees of nuclear installations regarding risk assessment and compliance with the safety criteria and define classification of licensing basis events. (author)

  18. Graphite selection for the PBMR reflector

    International Nuclear Information System (INIS)

    Marsden, B.J.; Preston, S.D.

    2000-01-01

    A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)

  19. The PBMR electric power generation plant

    International Nuclear Information System (INIS)

    Perez S, G.; Santacruz I, I.; Martin del Campo M, C.

    2003-01-01

    This work has as purpose to diffuse in a general way the technology of the one modulate reactor of pebble bed. Because our country is in developing ways, the electric power demand goes in increase with that which it is presented the great challenge of satisfying this necessity, not only being in charge of the one fact per se, but also involving the environmental aspect and of security. Both factors are covered by the PBMR technology, which we approach in their basic aspects with the purpose that the public opinion knows it and was familiarized with this type of reactors that well could represent a solution for our growing electricity demand. We will treat this reactor visualizing it like part of a generation plant defining in first place to the itself reactor. We will see because that the system PBMR consists of 2 main sections: the reactor and the unit of energy conversion, highlighting that the principle of the PBMR reactor operation is based on the thermodynamic Brayton cycle cooled by helium and that, in turn, it transmits the energy in form of heat toward a gas turbine. In what concerns to the fuel, it peculiar design due to its spherical geometry is described, aspect that make to this reactor different from the traditional ones that use fuel rods. In fact in the fuel spheres of the PBMR it is where it resides great part of it inherent security since each particle of fuel, consistent in uranium dioxide, is lined one with coal and silicon carbide those which form an impenetrable barrier containing to the fuel and those radioactive products that result of the nuclear reactions. Such particles are encapsulated in graphite to form the sphere or 'pebble', of here born the name of this innovative technology. (Author)

  20. The PBMR electric power generation plant; La planta de generacion de energia electrica PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Perez S, G.; Santacruz I, I.; Martin del Campo M, C. [FI-UNAM, 04500 Mexico D.F. (Mexico)] e-mail: gabriela_perez@engineer.com

    2003-07-01

    This work has as purpose to diffuse in a general way the technology of the one modulate reactor of pebble bed. Because our country is in developing ways, the electric power demand goes in increase with that which it is presented the great challenge of satisfying this necessity, not only being in charge of the one fact per se, but also involving the environmental aspect and of security. Both factors are covered by the PBMR technology, which we approach in their basic aspects with the purpose that the public opinion knows it and was familiarized with this type of reactors that well could represent a solution for our growing electricity demand. We will treat this reactor visualizing it like part of a generation plant defining in first place to the itself reactor. We will see because that the system PBMR consists of 2 main sections: the reactor and the unit of energy conversion, highlighting that the principle of the PBMR reactor operation is based on the thermodynamic Brayton cycle cooled by helium and that, in turn, it transmits the energy in form of heat toward a gas turbine. In what concerns to the fuel, it peculiar design due to its spherical geometry is described, aspect that make to this reactor different from the traditional ones that use fuel rods. In fact in the fuel spheres of the PBMR it is where it resides great part of it inherent security since each particle of fuel, consistent in uranium dioxide, is lined one with coal and silicon carbide those which form an impenetrable barrier containing to the fuel and those radioactive products that result of the nuclear reactions. Such particles are encapsulated in graphite to form the sphere or 'pebble', of here born the name of this innovative technology. (Author)

  1. Creation of the equilibrium core PBMR ORIGEN-S cross section library

    International Nuclear Information System (INIS)

    Stoker, C.C.; Reitsma, F.; Karriem, Z.

    2002-01-01

    As part of the design calculations for the Pebble Bed Modular Reactor (PBMR), fuel inventories, neutron and gamma sources and decay heat needs to be determined for the fuel spheres. Using the SCALE4.4 code system, a PBMR specific cross section library was created for the ORIGEN-S depletion calculations, assuming a 10-pass refueling system for the PBMR. In this paper the rationale for the creation of the PBMR library is evaluated in terms of the spectrum dependence due to burn-up. The ORIGEN-S PBMR library was further evaluated comparing the results for different parameters calculated with the reactor analysis diffusion code VSOP and the Monte Carlo code MCNP4C. (author)

  2. ZZ-PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design

    International Nuclear Information System (INIS)

    Reitsma, Frederik

    2007-01-01

    Description of benchmark: This international benchmark, concerns Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transients based on the PBMR-400 MW design. The deterministic neutronics, thermal-hydraulics and transient analysis tools and methods available to design and analyse PBMRs lag, in many cases, behind the state of the art compared to other reactor technologies. This has motivated the testing of existing methods for HTGRs but also the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for the design and safety evaluations of the PBMR. In addition to the development of new methods, this includes defining appropriate benchmarks to verify and validate the new methods in computer codes. The scope of the benchmark is to establish well-defined problems, based on a common given set of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark exercise has the following objectives: - Establish a standard benchmark for coupled codes (neutronics/thermal-hydraulics) for PBMR design; - Code-to-code comparison using a common cross section library ; - Obtain a detailed understanding of the events and the processes; - Benefit from different approaches, understanding limitations and approximations. Major Design and Operating Characteristics of the PBMR (PBMR Characteristic and Value): Installed thermal capacity: 400 MW(t); Installed electric capacity: 165 MW(e); Load following capability: 100-40-100%; Availability: ≥ 95%; Core configuration: Vertical with fixed centre graphite reflector; Fuel: TRISO ceramic coated U-235 in graphite spheres; Primary coolant: Helium; Primary coolant pressure: 9 MPa; Moderator: Graphite; Core outlet temperature: 900 C.; Core inlet temperature: 500 C.; Cycle type: Direct; Number of circuits: 1; Cycle

  3. Thermodynamic analysis of PBMR plant

    International Nuclear Information System (INIS)

    Sen, S.; Kadiroglu, O.K.

    2002-01-01

    The thermodynamic analysis of a PBMR is presented for various pressures and temperatures values. The design parameters of the components of the power plant are calculated and an optimum cycle for the maximum thermal efficiency is sought for. (author)

  4. Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

    This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

  5. Licensing of the proposed PBMR-SA

    International Nuclear Information System (INIS)

    Clapisson, G.A.; Henderson, N.R.; Hill, T.F.; Keenan, N.H.; Metcalf, P.E.; Mysenkov, A.

    1997-01-01

    This paper describes the preliminary criteria, which are intended to be used by the South African regulatory authority (CNS), for licensing of the ESKOM proposed - South African high temperature gas-cooled Pebble Bed Modular Reactor (PBMR-SA). The CNS intends to apply the existing CNS licensing approach together with some international design criteria used for advanced reactors as well as international experience gained from the safety evaluation of the MHTGR, THTR, etc. A major requirement to this type of reactor is that it should comply with the current CNS risk criteria and provide, as a minimum, the same degree of protection to the operator, public and environment that is required for the current generation of nuclear reactors. (author)

  6. PBMR desalination options: An economic study - HTR2008-58212

    International Nuclear Information System (INIS)

    De Bruyn, R.; Van Ravenswaay, J. P.; Hannink, R.; Kuhr, R.; Bhagat, K.; Zervos, N.

    2008-01-01

    The Pebble Bed Modular Reactor (PBMR), under development in South Africa, is an advanced helium-cooled graphite moderated high-temperature gas-cooled nuclear reactor. The heat output of the PBMR is primarily suited for process applications or power generation. In addition, various desalination technologies can be coupled to the PBMR to further improve the overall efficiency and economics, where suitable site opportunities exist. Several desalination application concepts were evaluated for both a cogeneration configuration as well as a waste heat utilization configuration. These options were evaluated to compare the relative economics of the different concepts and to determine the feasibility of each configuration. The cogeneration desalination configuration included multiple PBMR units producing steam for a power cycle, using a back-pressure steam turbine generator exhausting into different thermal desalination technologies. These technologies include Multi-Effect Distillation (MED), Multi-Effect Distillation with Thermal Vapor Compression (MED-TVC) as well as Multi-Stage Flash (MSF) with all making use of extraction steam from back-pressure turbines. These configurations are optimized to maximize gross revenue from combined power and desalinated water sales using representative economic assumptions with a sensitivity analysis to observe the impact of varying power and water costs. Increasing turbine back pressure results in a loss of power output but a gain in water production. The desalination systems are compared as incremental investments. A standard MED process with minimal effects appears most attractive, although results are very sensitive with regards to projected power and water values. (authors)

  7. The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

    International Nuclear Information System (INIS)

    Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

    2006-01-01

    The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

  8. Steady-state and accident analyses of PBMR with the computer code SPECTRA

    International Nuclear Information System (INIS)

    Stempniewicz, Marek M.

    2002-01-01

    The SPECTRA code is an accident analysis code developed at NRG. It is designed for thermal-hydraulic analyses of nuclear or conventional power plants. The code is capable of analysing the whole power plant, including reactor vessel, primary system, various control and safety systems, containment and reactor building. The aim of the work presented in this paper was to prepare a preliminary thermal-hydraulic model of PBMR for SPECTRA, and perform steady state and accident analyses. In order to assess SPECTRA capability to model the PBMR reactors, a model of the INCOGEN system has been prepared first. Steady state and accident scenarios were analyzed for INCOGEN configuration. Results were compared to the results obtained earlier with INAS and OCTOPUS/PANTHERMIX. A good agreement was obtained. Results of accident analyses with PBMR model showed qualitatively good results. It is concluded that SPECTRA is a suitable tool for analyzing High Temperature Reactors, such as INCOGEN or for example PBMR (Pebble Bed Modular Reactor). Analyses of INCOGEN and PBMR systems showed that in all analyzed cases the fuel temperatures remained within the acceptable limits. Consequently there is no danger of release of radioactivity to the environment. It may be concluded that those are promising designs for future safe industrial reactors. (author)

  9. Support to design and construction of the PBMR plant

    International Nuclear Information System (INIS)

    Cazorla, F.; Moron, P.; Gonzalez, J. I.

    2010-01-01

    Developing the new reactor design to a licensable state for constructing a pilot plant is a tough task require specific resources, concerning knowledge and previous experience, which trespassing the pure scientific or technologic knowledge linked to the reactor conceptual design. Taking into consideration the experience derived from the collaboration between the South African company PBMR (PTY) Ltd.; the Pebble Bed Modular Reactor Designer, and Tecnatom SA, the article presents some of the aspects in which the companies or organization in charge of the design can demand external support to license and construct the pilot plants with guaranteed success. (Author)

  10. Optimization of inlet plenum of A PBMR using surrogate modeling

    International Nuclear Information System (INIS)

    Lee, Sang-Moon; Kim, Kwang-Yong

    2009-01-01

    The purpose of present work is to optimize the design of inlet plenum of PBMR type gas cooled nuclear reactor numerically using a combining of three-dimensional Reynolds-averaged Navier-Stokes (RANS) analysis and surrogate modeling technique. Shear stress transport (SST) turbulence model is used as a turbulence closure. Three geometric design variables are selected, namely, rising channel diameter to plenum height ratio, aspect ratio of the plenum cross section, and inlet port angle. The objective function is defined as a linear combination of uniformity of three-dimensional flow distribution term and pressure drop in the inlet plenum and rising channels of PBMR term with a weighting factor. Twenty design points are selected using Latin-hypercube method of design of experiment and objective function values are obtained at each design point using RANS solver. (author)

  11. A Study for Burn-up Calculation applied on 400MWth PBMR Core

    International Nuclear Information System (INIS)

    Luu, Nam Hai; Kim, Hong Chul; Kim, Soon Young; Kim, Jong Kyung; Noh, Jae Man

    2007-01-01

    The 400MWth Pebble-bed Modular Reactor (PBMR) is an advanced high temperature gas cooled-reactor (HTGR). It possesses a very high efficiency and attractive economics without compromising the high levels of passive safety expected of advanced nuclear designs. With this reason, PBMR is a target which researchers especially in nuclear engineering field study carefully and therefore it is regarded as the leader in the power generation field. There are many research results about benchmark problems but results of the burn-up process are still poor. Hence, in this study a burn-up calculation was performed with PBMR using MONTEBURNS code in which MCNP modeling linked a depletion systems is used

  12. Reliability and integrity management program for PBMR helium pressure boundary components - HTR2008-58036

    International Nuclear Information System (INIS)

    Fleming, K. N.; Gamble, R.; Gosselin, S.; Fletcher, J.; Broom, N.

    2008-01-01

    The purpose of this paper is to present the results of a study to establish strategies for the reliability and integrity management (RIM) of passive metallic components for the PBMR. The RIM strategies investigated include design elements, leak detection and testing approaches, and non-destructive examinations. Specific combinations of strategies are determined to be necessary and sufficient to achieve target reliability goals for passive components. This study recommends a basis for the RIM program for the PBMR Demonstration Power Plant (DPP) and provides guidance for the development by the American Society of Mechanical Engineers (ASME) of RIM requirements for Modular High Temperature Gas-Cooled Reactors (MHRs). (authors)

  13. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-15

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  14. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    International Nuclear Information System (INIS)

    2013-04-01

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  15. Development and application of the PBMR fission product release calculation model

    International Nuclear Information System (INIS)

    Merwe, J.J. van der; Clifford, I.

    2008-01-01

    At PBMR, long-lived fission product release from spherical fuel spheres is calculated using the German legacy software product GETTER. GETTER is a good tool when performing calculations for fuel spheres under controlled operating conditions, including irradiation tests and post-irradiation heat-up experiments. It has proved itself as a versatile reactor analysis tool, but is rather cumbersome when used for accident and sensitivity analysis. Developments in depressurized loss of forced cooling (DLOFC) accident analysis using GETTER led to the creation of FIssion Product RElease under accident (X) conditions (FIPREX), and later FIPREX-GETTER. FIPREX-GETTER is designed as a wrapper around GETTER so that calculations can be carried out for large numbers of fuel spheres with design and operating parameters that can be stochastically varied. This allows full Monte Carlo sensitivity analyses to be performed for representative cores containing many fuel spheres. The development process and application of FIPREX-GETTER in reactor analysis at PBMR is explained and the requirements for future developments of the code are discussed. Results are presented for a sample PBMR core design under normal operating conditions as well as a suite of design-base accident events, illustrating the functionality of FIPREX-GETTER. Monte Carlo sensitivity analysis principles are explained and presented for each calculation type. The plan and current status of verification and validation (V and V) is described. This is an important and necessary process for all software and calculation model development at PBMR

  16. Preliminary safety analysis of a PBMR supplying process heat to a co-located ethylene production plant

    International Nuclear Information System (INIS)

    Scarlat, Raluca O.; Cisneros, Anselmo T.; Koutchesfahani, Tawni; Hong, Rada; Peterson, Per F.

    2012-01-01

    This paper considers the safety analysis and licensing approach for co-locating a pebble bed modular reactor (PBMR) to provide process heat to an ethylene production unit. The PBMR is an advanced nuclear reactor design that provides 400 MW of thermal energy. Ethylene production is an energy intensive process that utilizes large gas furnaces to provide the heat for the process. Coupling a PBMR with an ethylene production plant would open a new market for nuclear power, and would provide the chemical industry with a cleaner power source, helping to achieve the Clean Air Act standards, and eliminating the 0.5 ton of CO 2 emissions per ton of produced ethylene. Our analysis uses the Chevron Phillips chemical plant in Sweeney, TX as a prototypical site. The plant has four ethylene production trains, with a total power consumption of 2.4 GW, for an ethylene output of 3.7 million tons per year, 4% of the global ethylene production capacity. This paper proposes replacement of the gas furnaces by low-emission PBMR modules, and presents the safety concerns and risk mitigation and management options for this coupled system. Two coupling design options are proposed, and the necessary changes to the design basis events and severe accidents for the PBMR licensing application are discussed. A joint effort between the chemical and the nuclear entities to optimize the coupling design, establish preventive maintenance procedures, and develop emergency response plans for both of the units is recommended.

  17. Preliminary safety analysis of a PBMR supplying process heat to a co-located ethylene production plant

    Energy Technology Data Exchange (ETDEWEB)

    Scarlat, Raluca O., E-mail: rscarlat@nuc.berkeley.edu [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States); Cisneros, Anselmo T. [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States); Koutchesfahani, Tawni [University of California, Chemical and Biomolecular Engineering, 201 Gilman Hall, Berkeley, CA 94720 (United States); Hong, Rada; Peterson, Per F. [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States)

    2012-10-15

    This paper considers the safety analysis and licensing approach for co-locating a pebble bed modular reactor (PBMR) to provide process heat to an ethylene production unit. The PBMR is an advanced nuclear reactor design that provides 400 MW of thermal energy. Ethylene production is an energy intensive process that utilizes large gas furnaces to provide the heat for the process. Coupling a PBMR with an ethylene production plant would open a new market for nuclear power, and would provide the chemical industry with a cleaner power source, helping to achieve the Clean Air Act standards, and eliminating the 0.5 ton of CO{sub 2} emissions per ton of produced ethylene. Our analysis uses the Chevron Phillips chemical plant in Sweeney, TX as a prototypical site. The plant has four ethylene production trains, with a total power consumption of 2.4 GW, for an ethylene output of 3.7 million tons per year, 4% of the global ethylene production capacity. This paper proposes replacement of the gas furnaces by low-emission PBMR modules, and presents the safety concerns and risk mitigation and management options for this coupled system. Two coupling design options are proposed, and the necessary changes to the design basis events and severe accidents for the PBMR licensing application are discussed. A joint effort between the chemical and the nuclear entities to optimize the coupling design, establish preventive maintenance procedures, and develop emergency response plans for both of the units is recommended.

  18. Monte Carlo benchmark calculations for 400MWTH PBMR core

    International Nuclear Information System (INIS)

    Kim, H. C.; Kim, J. K.; Kim, S. Y.; Noh, J. M.

    2007-01-01

    A large interest in high-temperature gas-cooled reactors (HTGR) has been initiated in connection with hydrogen production in recent years. In this study, as a part of work for establishing Monte Carlo computation system for HTGR core analysis, some benchmark calculations for pebble-type HTGR were carried out using MCNP5 code. The core of the 400MW t h Pebble-bed Modular Reactor (PBMR) was selected as a benchmark model. Recently, the IAEA CRP5 neutronics and thermal-hydraulics benchmark problem was proposed for the testing of existing methods for HTGRs to analyze the neutronics and thermal-hydraulic behavior for the design and safety evaluations of the PBMR. This study deals with the neutronic benchmark problems, for fresh fuel and cold conditions (Case F-1), and first core loading with given number densities (Case F-2), proposed for PBMR. After the detailed MCNP modeling of the whole facility, benchmark calculations were performed. Spherical fuel region of a fuel pebble is divided into cubic lattice element in order to model a fuel pebble which contains, on average, 15000 CFPs (Coated Fuel Particles). Each element contains one CFP at its center. In this study, the side length of each cubic lattice element to have the same amount of fuel was calculated to be 0.1635 cm. The remaining volume of each lattice element was filled with graphite. All of different 5 concentric shells of CFP were modeled. The PBMR annular core consists of approximately 452000 pebbles in the benchmark problems. In Case F-1 where the core was filled with only fresh fuel pebble, a BCC(body-centered-cubic) lattice model was employed in order to achieve the random packing core with the packing fraction of 0.61. The BCC lattice was also employed with the size of the moderator pebble increased in a manner that reproduces the specified F/M ratio of 1:2 while preserving the packing fraction of 0.61 in Case F-2. The calculations were pursued with ENDF/B-VI cross-section library and used sab2002 S(α,

  19. Verification of Serpent code for the fuel analysis of a PBMR; Verificacion del codigo SERPENT para el analisis de combustible para un PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the models and simulations with the Monte Carlo code Serpent are presented, as well as the obtained results of the different analyzed cases in order to verify the suitability or reliability of the use of this code to ensure favorable results in the realization of a neutronic analysis of fuel for a Pebble Bed Modular Reactor (PBMR). Comparisons were made with the results reported in a report by the OECD/Nea relative to a high temperature reactor of spheres bed with plutonium reactor grade as fuel. The results show that the use of Serpent is appropriate, as these results are comparable with those reported in the report. (Author)

  20. Analysis of Dust and Fission Products in PBMR Turbine

    International Nuclear Information System (INIS)

    Stempniewicz, M.M.; Wessels, D.

    2014-01-01

    A 400 MWth direct cycle Pebble Bed Modular reactor was under development in South Africa. The work performed included design and safety analyses. In HTR/PBMR, graphite dust is generated during normal reactor operation due to pebble-to-pebble scratching. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released, although in very small quantities, during normal operation. This paper presents a model and analyses of the PBMR turbine with the SPECTRA code. The purpose of the present work was to estimate the amount and distribution of deposited dust and the fission products, namely cesium, iodine, and silver, during plant life-time, which was assumed to be 40 full-power years. The performed work showed that after 40 years of plant life-time deposited layers are very small. The largest deposition is of course observed on the dust filters. Apart from the dust filters, the largest dust deposition is observed on the: • Outer Casing (inner walls) • Turbine Rotor Cooling Cavity (inner walls) • HPC Cold Cooling Gas Header (inner walls) This is caused by relatively low gas velocities in these volumes. The low velocities allow a continuous build-up of the dust layer. About 90% of cesium, 40% of iodine, and 99.9% of silver is adsorbed on the metallic structures of the turbine. The sorption rate increases along the turbine due to decreasing temperatures. In case of cesium and iodine the highest concentrations are observed in the last stage (stage 12) of the turbine. In the case of silver the sorption is so large that the silver vapor is significantly depleted in the last stages of the turbine. This is a reason for having a maximum in silver concentration in the stage 10. In the following stages the concentration decreases due to very small silver vapor fraction in the gas. (author)

  1. Xenon-induced axial power oscillations in the 400 MW PBMR

    International Nuclear Information System (INIS)

    Strydom, Gerhard

    2008-01-01

    The redistribution of the spatial xenon concentration in the 400 MW Pebble Bed Modular Reactor (PBMR) core has a non-linear, time-dependent feedback effect on the spatial power density during several types of operational transient events. Due to the inherent weak coupling that exists between the iodine and xenon formation and destruction rates, as well as the complicating effect of spatial variance in the thermal flux field, reactor cores have been analyzed for a number of decades for the occurrence and severity of xenon-induced axial power oscillations. Of specific importance is the degree of oscillation damping exhibited by the core during transients, which involves axial variations in the local power density. In this paper the TINTE reactor dynamics code is used to assess the stability of the current 400 MW PBMR core design with regard to axial xenon oscillations. The focus is mainly on the determination of the inherent xenon and power oscillation damping properties by utilizing a set of hypothetical control rod insertion transients at various power levels. The oscillation damping properties of two 100%-50%-100% load-follow transients, one of which includes the de-stabilizing axial effects of moving control rods, are also discussed in some detail. The study shows that, although first axial mode oscillations do occur in the 400 MW PBMR core, the inherent damping of these oscillations is high, and that none of the investigated load-follow transients resulted in diverging oscillations. It is also shown that the PBMR core exhibits no radial oscillation components for these xenon-induced axial power oscillations

  2. Verification of Serpent code for the fuel analysis of a PBMR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2015-09-01

    In this paper the models and simulations with the Monte Carlo code Serpent are presented, as well as the obtained results of the different analyzed cases in order to verify the suitability or reliability of the use of this code to ensure favorable results in the realization of a neutronic analysis of fuel for a Pebble Bed Modular Reactor (PBMR). Comparisons were made with the results reported in a report by the OECD/Nea relative to a high temperature reactor of spheres bed with plutonium reactor grade as fuel. The results show that the use of Serpent is appropriate, as these results are comparable with those reported in the report. (Author)

  3. Evolution of near term PBMR steam and cogeneration applications - HTR2008-58219

    International Nuclear Information System (INIS)

    Kuhr, R. W.; Hannink, R.; Paul, K.; Kriel, W.; Greyvenstein, R.; Young, R.

    2008-01-01

    US and international applications for large onsite cogeneration (steam and power) systems are emerging as a near term market for the PBMR. The South African PBMR demonstration project applies a high temperature (900 deg. C) Brayton cycle for high efficiency power generation. In addition, a number of new applications are being investigated using an intermediate temperature range (700-750 deg. C) with a simplified heat supply system design. This intermediate helium delivery temperature supports conventional steam Rankine cycle designs at higher efficiencies than obtained from water type reactor systems. These designs can be adapted for cogeneration of steam, similar to the design of gas turbine cogeneration plants that supply steam and power at many industrial sites. This temperature range allows use of conventional or readily qualifiable materials and equipment, avoiding some cost premiums associated with more difficult operating conditions. As gas prices and CO 2 values increase, the potential value of a small nuclear reactor with advanced safety characteristics increases dramatically. Because of its smaller scale, the 400-500 MWt PBMR offers the economic advantages of onsite thermal integration (steam, hot water and desalination co-production) and of providing onsite power at cost versus at retail industrial rates avoiding transmission and distribution costs. Advanced safety characteristics of the PBMR support the location of plants adjacent to steam users, district energy systems, desalination plants, and other large commercial and industrial facilities. Additional benefits include price stability, long term security of energy supply and substantial CO 2 reductions. Target markets include existing sites using gas fired boilers and cogeneration units, new projects such as refinery and petrochemical expansions, and coal-to-liquids projects where steam and power represent major burdens on fuel use and CO 2 emissions. Lead times associated with the nuclear licensing

  4. Challenges and opportunities in providing a digital protection system for the PBMR - HTR2008-58173

    International Nuclear Information System (INIS)

    Marais, J.; Ridolfo, C. F.

    2008-01-01

    The Republic of South Africa is currently developing the Pebble Bed Modular Reactor (PBMR); an advanced, fourth-generation reactor that incorporates inherent safety features, which require no human intervention and which provide an unprecedented level of nuclear safety. In addition to electrical power generation, the reactor is uniquely suited for a variety of non-traditional nuclear applications including oil sands extraction, desalination, and hydrogen production. A state-of-the-art digital Protection System for the PBMR is currently being developed in conjunction with Westinghouse Electric Company (WEC). The Protection System provides for: reactor shutdown using two different reactor trip methodologies (dropping of the control rods and insertion of Small Absorber Spheres (SASs) which are composed of boron carbide); post-event monitoring; and manual reactor shutdown, which is independent of software-based systems. The reactor shutdown and post-event instrumentation monitoring components of the Protection System are being implemented utilizing the WEC 'Common Q' platform, which is comprised of 'commercially dedicated' Programmable Logic Controllers (PLCs), colour-graphic Flat Panel Displays (FPDs) with integral touch screens, and high-speed data communication links. High reliability and availability are achieved through component redundancy, continuous automatic self-testing which is run online in a background mode, and implementation of a multi-channel system design which is tolerant to failures. The Protection System is also designed to support periodic surveillance testing through a suite of built- in computer-aided test facilities that are accessible via an FPD interface. These allow various system surveillance requirements to be readily performed in a convenient and systematic manner. This paper discusses the following topics with regard to the PBMR Protection System: development strategy, functional requirements, selection of applicable Codes and Standards

  5. Fuel analysis of a PBMR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2015-09-01

    In this paper a neutronic analysis of fuel for a Pebble Bed Modular Reactor is presented, based on their composition and geometric distribution, having as main objective the use and utilization of thorium for the production of fuel for the operation of this reactor. For the study of these characteristics is necessary to use a code capable of carry out a reliable calculation of the main parameters of the fuel. Using the Monte Carlo method is suitable for simulating the neutron transport in the reactor core, which is the basis of Serpent code, with which the calculations for the analysis will be made. The results show the desirability of the use of thorium, since presents good conversion levels of fertile material to fissile, to produce U 233 by neutron capture, taking as a very important factor the distribution of materials in the core, which in this work had better results based on the neutron multiplication effective factor, formed by three right circular cylinders circumscribed, making that the core has three areas constituted by a mixture of plutonium oxide in the central and external areas, and thorium oxide in the intermediate area. (Author)

  6. Development of a strategy for the management of PBMR spent fuel in South Africa

    Energy Technology Data Exchange (ETDEWEB)

    Smith, S.W., E-mail: Schalk.Smith@necsa.co.z [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa); Bredell, P.J. [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa)

    2010-10-15

    South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would form an integral part of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development.

  7. Test Results of PBMR Fuel Spheres

    International Nuclear Information System (INIS)

    Koshcheev, Konstantin; Diakov, Alexander; Beltyukov, Igor; Barybin, Andrey; Chernetsov, Mikhail

    2014-01-01

    Results of pre-irradiation testing of fuel spheres (FS) and coated particles (CP) manufactured by PBMR SOC (Republic of South Africa) are described. The stable high quality level of major characteristics (dimensions, CP coating structure, uranium-235 contamination of the FS matrix graphite and the outer PyC layer of the CP coating) are shown. Results of a methodical irradiation test of two FS in helium and neon medium at temperatures of 800 to 1300 °C with simultaneous determination of release-to-birth ratios for major gaseous fission products (GFP) are described. (author)

  8. A study on the design concepts of the PBMR and the GT-MHR

    International Nuclear Information System (INIS)

    Park, Won Seok; Chang, Jong Hwa; Park, Chang Kue

    2004-05-01

    The major application of the nuclear power in the energy sector has been to produce the electricity. However, a growing concern on the environment and the expected shortage of the fossil energy resources is demanding the expansion of nuclear energy's role in the energy sectors. The High Temperature Gas cooled Reactor (HTGR) has been expected to expand the role of nuclear energy because of its high temperature capability. Especially, the interest on the HTGR has been sharply increased recently related with the production of the hydrogen. About 5 HTGRs had been operated by the end of 1980s. However, all of them had been terminated permanently at the end of 1980s because of their poor system economy and frequent technical troubles. A new concept called MHTGR (Modular High Temperature Gas cooled Reactor) emerged in early of 1990s. Two MHTGR concepts on commercial basis have been developed since then; one is the PBMR (Pebble Bed Modular Reactor) developed by Eskom in South Africa and another is GT-MHR (Gas Turbine Modular High-temperature Reactor) developed by both GA in USA and OKBM in Russia. In this report, the design concepts for the PBMR and GT-MHR were reviewed

  9. A study on the design concepts of the PBMR and the GT-MHR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Won Seok; Chang, Jong Hwa; Park, Chang Kue

    2004-05-01

    The major application of the nuclear power in the energy sector has been to produce the electricity. However, a growing concern on the environment and the expected shortage of the fossil energy resources is demanding the expansion of nuclear energy's role in the energy sectors. The High Temperature Gas cooled Reactor (HTGR) has been expected to expand the role of nuclear energy because of its high temperature capability. Especially, the interest on the HTGR has been sharply increased recently related with the production of the hydrogen. About 5 HTGRs had been operated by the end of 1980s. However, all of them had been terminated permanently at the end of 1980s because of their poor system economy and frequent technical troubles. A new concept called MHTGR (Modular High Temperature Gas cooled Reactor) emerged in early of 1990s. Two MHTGR concepts on commercial basis have been developed since then; one is the PBMR (Pebble Bed Modular Reactor) developed by Eskom in South Africa and another is GT-MHR (Gas Turbine Modular High-temperature Reactor) developed by both GA in USA and OKBM in Russia. In this report, the design concepts for the PBMR and GT-MHR were reviewed.

  10. The first stage of licensing of PBMR in South Africa and safety issues

    International Nuclear Information System (INIS)

    Clapisson, G.A.; Mysen, A.

    2002-01-01

    The National Nuclear Regulator (NNR) has received a nuclear installation licence application from Eskom (the South African electricity utility). The Application is made in accordance with the National Nuclear Regulator Act for a nuclear installation licence for the demonstration module of a 110 MWe Class Pebble Bed Modular Reactor (PBMR) electricity generating power station. It is proposed to locate the installation on Eskom property within the owner-controlled boundary of Koeberg Nuclear Power Station situated in the Western Cape, subject to inter alia a favourable Environmental Impact Assessment (EIA) record of decision, which is currently being undertaken under the requirements of another legislation the Environment Conservation Act. The PBMR is a graphite moderated helium cooled reactor using a direct gas cycle to convert the heat, generated by nuclear fission in the reactor and transferred to the coolant gas, into electrical energy by means of a helium turbo-generator. By design, provision has been made to accommodate the storage of spent fuel in the buildings for the 40-year design life of the plant and thereafter for a further period if so required. Radioactive material and waste will be managed and disposed of in accordance with Regulatory and Government legal requirements. (authors)

  11. Helium storage and control system for the PBMR

    International Nuclear Information System (INIS)

    Verkerk, E.C.

    1997-01-01

    The power conversion unit will convert the heat energy in the reactor core to electrical power. The direct-closed cycle recuperated Brayton Cycle employed for this concept consists of a primary helium cycle with helium powered turbo compressors and a power turbine. The helium is actively cooled with water before the compression stages. A recuperator is used to preheat the helium before entering the core. The start of the direct cycle is initiated by a mass flow from the helium inventory and control system via a jet pump. When the PBMR is connected to the grid, changes in power demand can be followed by changing the helium flow and pressure inside the primary loop. Small rapid adjustments can be performed without changing the helium inventory of the primary loop. The stator blade settings on the turbines and compressors are adjustable and it is possible to bypass reactor and turbine. This temporarily reduces the efficiency at which the power conversion unit is operating. Larger or long term adjustments require storage or addition of helium in order to maintain a sufficient level of efficiency in the power conversion unit. The helium will be temporarily stored in high pressure tanks. After a rise in power demand it will be injected back into the system. Some possibilities how to store the helium are presented in this paper. The change of helium inventory will cause transients in the primary helium loop in order to acquire the desired power level. At this stage, it seems that the change of helium inventory does not strongly effect the stability of the power conversion unit. (author)

  12. Neutronic analysis of the PBMR-400 full core using thorium fuel mixed with plutonium or minor actinides

    International Nuclear Information System (INIS)

    Acır, Adem; Coşkun, Hasan

    2012-01-01

    Highlights: ► Neutronic calculations for PBMR 400 were conducted with the computer codes MCNP and MONTEBURNS 2.0. ► The criticality and burnup were investigated for reactor grade plutonium and minor actinides. ► We found that the use of these new fuels in PBMRs would reduce the nuclear waste repository significantly. -- Abstract: Time evolution of criticality and burnup grades of the PBMR were investigated for reactor grade plutonium and minor actinides in the spent fuel of light water reactors (LWRs) mixed with thoria. The calculations were performed by employing the computer codes MCNP and MONTEBURNS 2.0 and using the ENDF/B-V nuclear data library. Firstly, the plutonium–thorium and minor actinides–thorium ratio was determined by using the initial k eff value of the original uranium fuel design. After the selection of the plutonium/minor actinides–thorium mixture ratio, the time-dependent neutronic behavior of the reactor grade plutonium and minor actinides and original fuels in a PBMR-400 reactor was calculated by using the MCNP code. Finally, k eff , burnup and operation time values of the fuels were compared. The core effective multiplication factor (k eff ) for the original fuel which has 9.6 wt.% enriched uranium was computed as 1.2395. Corresponding to this k eff value the reactor grade plutonium/thorium and minor actinide/thorium oxide mixtures were found to be 30%/70% and 50%/50%, respectively. The core lives for the original, the reactor grade plutonium/thorium and the minor actinide/thorium fuels were calculated as ∼3.2, ∼6.5 and ∼5.5 years, whereas, the corresponding burnups came out to be 99,000, ∼190,000 and ∼166,000 MWD/T, respectively, for an end of life k eff set equal to 1.02.

  13. The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

    International Nuclear Information System (INIS)

    Pautz, A.; Tyobeka, B.; Ivanov, K.

    2009-01-01

    In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

  14. Development of a strategy for the management of PBMR spent fuel in South Africa - HTR2008-58047

    International Nuclear Information System (INIS)

    Smith, S. W.; Bredell, P. J.; Meyer, W. C. M. H.

    2008-01-01

    South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development. (authors)

  15. The OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark - Steady-state results and status

    International Nuclear Information System (INIS)

    Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.

    2008-01-01

    The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)

  16. Separate effects tests to determine the effective thermal conductivity in the PBMR HTTU test facility

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, P.G., E-mail: pgr@mtechindustrial.com [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X6001, Potchefstroom 2520 (South Africa); Toit, C.G. du; Antwerpen, W. van [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X6001, Potchefstroom 2520 (South Africa); Antwerpen, H.J. van [M-Tech Industrial (Pty) Ltd., PO Box 19855, Noordbrug 2522 (South Africa)

    2014-05-01

    Thermal-fluid simulations are used extensively to predict the maximum fuel temperatures, flows, pressure drops and thermal capacitance of pebble bed gas cooled reactors in support of the reactor safety case. The PBMR company developed the HTTU non-nuclear test facility in cooperation with M-Tech Industrial (Pty) Ltd. and the North-West University in South Africa to conduct comprehensive separate effects tests as well as integrated effects tests to study the different thermal-fluid phenomena. This paper describes the separate effects tests that were conducted to determine the effective thermal conductivity through the pebble bed under near-vacuum conditions and temperatures up to 1200 °C. It also presents the measured temperature distributions and the methodology applied in the data analysis to derive the resultant values of effective thermal conductivity and its associated uncertainty.

  17. Separate effects tests to determine the effective thermal conductivity in the PBMR HTTU test facility

    International Nuclear Information System (INIS)

    Rousseau, P.G.; Toit, C.G. du; Antwerpen, W. van; Antwerpen, H.J. van

    2014-01-01

    Thermal-fluid simulations are used extensively to predict the maximum fuel temperatures, flows, pressure drops and thermal capacitance of pebble bed gas cooled reactors in support of the reactor safety case. The PBMR company developed the HTTU non-nuclear test facility in cooperation with M-Tech Industrial (Pty) Ltd. and the North-West University in South Africa to conduct comprehensive separate effects tests as well as integrated effects tests to study the different thermal-fluid phenomena. This paper describes the separate effects tests that were conducted to determine the effective thermal conductivity through the pebble bed under near-vacuum conditions and temperatures up to 1200 °C. It also presents the measured temperature distributions and the methodology applied in the data analysis to derive the resultant values of effective thermal conductivity and its associated uncertainty

  18. DORT-TD/THERMIX solutions for the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

    International Nuclear Information System (INIS)

    Tyobeka, Bismark; Pautz, Andreas; Ivanov, Kostadin

    2008-01-01

    In new reactor designs that are still under review such as the PBMR, not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400 MW OECD/NEA/NSC coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate the transient scenarios in the above-mentioned benchmark problem. Steady-state calculations results are compared with selected participants' results as well as transient models in which the diffusion and transport theory solutions of the same code system are directly compared. Several sensitivity studies are also shown in order to determine how much the change in certain parameters influences the overall behaviour of a given transient. It is shown in this paper that DORT-TD/THERMIX is a versatile tool which can be deployed for design and safety analyses of high temperature reactors of pebble-bed type. (authors)

  19. The ESKOM pebble bed modular reactor

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1999-01-01

    An audit has been made of the design, construction, safety, economics and marketability of the ESKOM pebble bed modular reactor (PBMR). In this paper that audit is briefly summarized. The principal conclusions of the audit are as follows. The design is sound. It is a logical development of the designs proposed for other, modern, high-temperature gas-cooled reactors. More than 80% of the cost of constructing and commissioning a series of PBMRs would be spent in South Africa. The PBMR is much safer than existing nuclear power reactors and for many practical purposes it may be treated as a conventional chemical plant. The PBMR is economically competitive with thermal power stations. There is a substantial global market for the PBMR. (author)

  20. Analysis of PBMR transients using a coupled neutron transport/thermal-hydraulics code DORT-TD/thermix

    International Nuclear Information System (INIS)

    Tyobeka, B.; Ivanov, K.; Pautz, A.

    2007-01-01

    In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters

  1. Introduction to the PBMR heat transfer test facility

    International Nuclear Information System (INIS)

    Rousseau, P.G.; Staden, M. van

    2008-01-01

    This paper provides an introduction to the Heat Transfer Test Facility (HTTF) that is currently being developed for PBMR (Pty.) Ltd. by M-Tech Industrial (Pty.) Ltd. in association with North-West University in South Africa. The paper provides an overview of the phenomena that will be studied and the envisaged test configurations for each of these phenomena. It also shows the layouts of the different test units namely the High Pressure Test Unit (HPTU) and the High Temperature Test Unit (HTTU) and provides an overview of the planned test schedule

  2. The OECD/NEA/NSC PBMR 400 MW coupled neutronics thermal hydraulics transient benchmark: transient results - 290

    International Nuclear Information System (INIS)

    Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.

    2010-01-01

    The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)

  3. Separate effects tests to determine the thermal dispersion in structured pebble beds in the PBMR HPTU test facility

    Energy Technology Data Exchange (ETDEWEB)

    Toit, C.G. du, E-mail: jat.dutoit@nwu.ac.za; Rousseau, P.G.; Kgame, T.L.

    2014-05-01

    Thermal-fluid simulations are used extensively to predict the maximum fuel temperatures, flows, pressure drops and thermal capacitance of pebble bed gas cooled reactors in support of the reactor safety case. The PBMR company developed the HTTF test facility in cooperation with M-Tech Industrial (Pty) Ltd. and the North-West University in South Africa to conduct comprehensive separate effects tests as well as integrated effects tests to study the different thermal-fluid phenomena. This paper describes the separate effects tests that were conducted to determine the effect of the porous structure on the fluid effective thermal conductivity due to the thermal dispersion. It also presents the methodology applied in the data analysis to derive the resultant values of the effective thermal conductivity and its associated uncertainty.

  4. Design and simulation of a process of seawater desalination (MED) using the residual heat of a PBMR nuclear power plant; Diseno y simulacion de un proceso de desalinizacion de agua de mar (MED) utilizando el calor residual de una planta nucleoelectrica PBMR

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

    In the present work it is demonstrated as the thermodynamic recuperative Brayton cycle with which operates a nuclear power plant type PBMR (Pebble Bed Modular Reactor) it allows to use the residual heat, removed in the coolers of the compression stage of the system, to produce vapor and to desalt seawater. The desalination process selected, starting from its operation characteristics and the derived advantages of them using nuclear heat, it the one of Multi-Effect Distillation, MED for its abbreviations in English, which described and it is justified to detail. This distillation process widely studied, allows us to use water vapor pressurized to temperatures between 70 and 110 C like energy source to evaporate the seawater in the first stage or effect of the process. The relatively low temperatures with which the vapor takes place of feeding to the process is it makes to the plant PBMR ideal for desalination of seawater, since does not require majors modifications to its design its operation, and on the contrary it allows to use the heat that previously was rejected, to produce the vapor. In this work an unit MED of six effects is designed, which undergoes a successive fall of pressure in each of them. Once obtained the agreed design to the conditions of operation of PBMR plant, it was model mathematically the MED process, including the coupling stage with the reactor coolers. The mathematical model was obtained by means of differential equations of mass balance and energy in the system, and with this it was implemented in SIMULINK a model equivalent to the MED process which is interconnected to the simulator coolers of the PBMR plant, constructed previously. One ran the program being obtained the results that are reported at the end of this article. (Author)

  5. The PBMR fuel plant: Proven technology in an advanced safety environment

    International Nuclear Information System (INIS)

    Braehler, G.; Froschauer, K.; Welbers, P.; Boyes, D.

    2008-01-01

    The PBMR Fuel Plant (PFP), to be constructed at the Pelindaba site near Johannesburg will fuel the first South African Pebble Bed Modular Reactor. The qualification of the PBMR fuel shall be based on past experience with fuel which was produced in the German NUKEM/HOBEG plant and irradiated in the German AVR reactor. Accordingly, the PFP must produce the same fuel as the German plant did, and consequently, the design of the PFP has in essence to be a copy of the NUKEM/HOBEG plant. As a reminder this plant had been operated in accordance with the German regulatory rules which were defined in the years 1970/80. Since then, the requirements with regard to radiological protection, criticality safety and emission control have been significantly tightened, and of course the PFP must be designed in accordance with the most advanced international norms and standards. The implications which follow from these two potentially conflicting requirements, as defined above, are highlighted, and technical solutions are presented. Hence, the change from administrative criticality safety control to technical control, i.e. the application of safe geometry as far as possible. and the introduction of technical solutions for the remaining safe mass regime will be described. A lot of equipment in the Kernel area and in the recycling areas needed to be redesigned in safe geometry. The sensitive processes for Kernel Calcining, for the Coating and the Over-coating remain under safe mass regime, but the safety against criticality is completely independent from staff activities and based on technical measures. A new concept for safe storage of large volumes of Uranium-containing liquids has been developed. Also, the change from relatively open handling of Uranium to the application of containment enclosures wherever release of radioactivity into the room atmosphere is possible, will be addressed. This change required redesign of all process steps requiring the handling of dry Uranium oxides

  6. Fuel analysis of a PBMR; Analisis de combustible de un PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper a neutronic analysis of fuel for a Pebble Bed Modular Reactor is presented, based on their composition and geometric distribution, having as main objective the use and utilization of thorium for the production of fuel for the operation of this reactor. For the study of these characteristics is necessary to use a code capable of carry out a reliable calculation of the main parameters of the fuel. Using the Monte Carlo method is suitable for simulating the neutron transport in the reactor core, which is the basis of Serpent code, with which the calculations for the analysis will be made. The results show the desirability of the use of thorium, since presents good conversion levels of fertile material to fissile, to produce U{sup 233} by neutron capture, taking as a very important factor the distribution of materials in the core, which in this work had better results based on the neutron multiplication effective factor, formed by three right circular cylinders circumscribed, making that the core has three areas constituted by a mixture of plutonium oxide in the central and external areas, and thorium oxide in the intermediate area. (Author)

  7. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Brian David [Los Alamos National Laboratory; Beddingfield, David H [Los Alamos National Laboratory; Durst, Philip [INL; Bean, Robert [INL

    2010-01-01

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

  8. Aprendizaje basado en juegos

    Directory of Open Access Journals (Sweden)

    Marco A. Gómez-Martín

    2012-04-01

    Full Text Available Gracias al incremento de potencia de los ordenadores, gran cantidad de personas dedican horas y horas a aprovechar su aspecto más lúdico, los videojuegos. Por otro lado, existen programas educativos que aprovechan la infinita paciencia de los ordenadores que les hacen capaces de explicar conceptos una y otra vez hasta que los alumnos lo entiendan. En este artículo mostramos qué cosas pueden aportar las aplicaciones de enseñanza a los videojuegos y viceversa. Terminamos describiendo JV2M, como un ejemplo de sistema de aprendizaje basado en juegos.

  9. Generation 4 - nuclear reactors and an approach to secure public acceptance and access to energy for everyone

    International Nuclear Information System (INIS)

    Pahladsingh, R.

    2001-01-01

    The aim of this paper is to bring the Pebble Bed Modular Reactor (PBMR) and a few interesting Light Water Passive nuclear reactor designs under your attention. The PBMR is under further development in South Africa and Asia. The philosophy behind the PBMR concept has been to develop a nuclear reactor which is so safe that it could be called inherently safe. Its concept is so completely different, see figure 2, that it can easily pass strictest safety regulations. Consequently it is a good Generation IV candidate. Good promotion of the gas-turbine direct cycle PBMR design is a main task to the nuclear technology and industry and could be the challenge that the young generation needs to consider a career in nuclear technology. (authors)

  10. Design of a physical model of the PBMR with the aid of Flownet

    International Nuclear Information System (INIS)

    Greyvenstein, G.P.; Rousseau, P.G.

    2002-01-01

    The design of a physical model of the PBMR with the aid of the code Flownet is discussed in this paper. The purpose of the physical model is to test the control strategies and operating procedures of the PBMR and also to demonstrate the accuracy of Flownet. Flownet is first used to do component matching and to determine the detail steady-state performance of the system. It is then demonstrated how the code was used to simulate the start-up procedure as well as a load following and a load rejection scenario. The study demonstrates how a micro model of the PBMR can be designed with the aid of a powerful simulation tool in a relatively short period of time and at low cost using commercially available turbochargers. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  12. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-08-15

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

  14. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    2011-08-01

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

  15. A multi-tank storage facility to effect power control in the PBMR power cycle

    International Nuclear Information System (INIS)

    Matimba, T.A.D.; Krueger, D.L.W.; Mathews, E.H.

    2007-01-01

    This article presents the concept of a storage facility used to effect power control in South Africa's PBMR power cycle. The concept features a multiple number of storage vessels whose purpose is to contain the working medium, helium, as it is withdrawn from the PBMR's closed loop power cycle, at low energy demand. This helium is appropriately replenished to the power cycle as the energy demand increases. Helium mass transfer between the power cycle and the storage facility, henceforth known as the inventory control system (ICS), is carried out by way of the pressure differential that exists between these two systems. In presenting the ICS concept, emphasis is placed on storage effectiveness; hence the discussion in this paper is centred on those features which accentuate storage effectiveness, namely:- Storage vessel multiplicity; - Unique initial pressures for each vessel arranged in a cascaded manner; and - A heat sink placed in each vessel to provide thermal inertia. Having presented the concept, the objective is to qualitatively justify the presence of each of the above-mentioned features using thermodynamics as a basis

  16. PBMR spent fuel bulk dry storage heat removal - HTR2008-58170

    International Nuclear Information System (INIS)

    De Wet, G. J.; Dent, C.

    2008-01-01

    A low decay heat (implying Spent Fuel (SF) pebbles older than 8-9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks' vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading/unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell. (authors)

  17. Integrated design approach of the pebble BeD modular reactor using models

    International Nuclear Information System (INIS)

    Venter, Pieter J.; Mitchell, Mark N.

    2007-01-01

    The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR

  18. The Pebble Bed Modular Reactor: An obituary

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Steve, E-mail: stephen.thomas@gre.ac.u [Public Services International Research Unit (PSIRU), Business School, University of Greenwich, 30 Park Row, London SE10 9LS (United Kingdom)

    2011-05-15

    The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money. - Research highlights: {yields} In this study we examine the reasons behind the failure of the South African PBMR programme. {yields} The study reviews the technical issues that have arisen and lessons for future reactor developments. {yields} The study also identifies weaknesses in the accountability mechanisms for public spending.

  19. From field to factory-Taking advantage of shop manufacturing for the pebble bed modular reactor

    International Nuclear Information System (INIS)

    Wallace, Edward; Matzie, Regis; Heiderd, Roger; Maddalena, John

    2006-01-01

    The move of nuclear plant construction from the field to the factory for small, advanced pebble bed modular reactor (PBMR) designs has significant benefits compared to traditional light water reactor (LWR) field oriented designs. The use of modular factory construction techniques has a growing economic benefit over time through well-established process learning applications. This paper addresses the basic PBMR design objectives and commercialization model that drive this approach; provides a brief technical description of the PBMR design and layout with representative CAD views and discusses derived figures of merit highlighting the relative simplicity of PBMR compared to a modern LWR. The discussion emphasizes that more of PBMR can be built in the factory due to the simple design of a direct helium Brayton cycle compared to an indirect LWR steam cycle with its associated equipment. For the PBMR design there are fewer and less cumbersome auxiliary and safety systems with their attendant support requirements. Additionally, the labor force economic efficiency for nuclear projects is better in the factory than in the field, including consideration of labor costs and nuclear quality programs. Industrial learning is better in the factory because of the more controlled environment, mechanization optimization opportunities and because of the more stable labor force compared to the field. Supply chain benefits are more readily achievable with strategic contracts for module suppliers. Although building a nuclear power plant is not a typical high volume manufacturing process, for the PBMR-type of plant, with its high degree of standardization and relatively small, simplified design, the shift to factory work has a significant impact on overall project cost due to earlier identification and better coordination of parallel construction paths. This is in stark contrast to the construction of a large LWR in the past. Finally, the PBMR modular plant concept continues at the

  20. South Africa's nuclear model: A small and innovative reactor is seen as the model for new electricity plants. The project is nearing the starting blocks

    International Nuclear Information System (INIS)

    Ferreira, Tom

    2004-01-01

    Although nuclear power generation has by far the best safety and environmental record of any technology in general use, it has for many years been unable to make any meaningful inroads into the wall of negative perceptions that have arisen against it. But sentiments are changing rapidly on a global scale. The flare-up of oil prices is a sobering reminder of the volatility in the energy market, the exhaustibility of fossil fuels and the urgent need for stable, reliable, non-polluting sources of electrical power that are indispensable to a modern industrial economy. Today, new types of nuclear plants are prized, and South Africa is moving ahead. The State energy provider, Eskom, is internationally regarded as the leader in the field of the Pebble Bed Modular Reactor (PBMR) technology, a 'new generation' nuclear power plant. A decision on the PBMR project's future is on the near horizon. Should approvals be received in the coming months to proceed to the project's next phase, construction of the PBMR demonstration plant will start in 2006, in which case the reactor will start in 2010 and handed over to the client, Eskom, in 2011. Eskom has conditionally undertaken to purchase the first commercial units. Pebble bed reactors are small, about one-sixth the size of most current nuclear plants. Multiple PBMRs can share a common control center and occupy an area of no more than three football fields. More specifically, the PBMR is a helium-cooled, graphite moderated high temperature reactor (HTR). The concept is based on experience in the UK, United States and particularly Germany where prototype reactors were operated successfully between the late 1960s and 1980s. Although it is not the only high-temperature, gas-cooled nuclear reactor being developed in the world, the South African project is internationally regarded as a front-runner. The South African PBMR includes unique and patented technological innovations which make it particularly competitive. The Chief Executive

  1. Effect of high temperature annealing on the grain size of CVD-grown SiC and experimental PBMR TRISO coated particles

    CSIR Research Space (South Africa)

    Mokoduwe, SM

    2010-10-01

    Full Text Available in the PBMR fuel SiC layer. square samples were cut from the original sample received from ORNL and prepared for grain size Prague, Czech Republic, October 18 – 2000 °C. These no significant ion of how the 8] also ge is also of tal THODS -Si... for grain size determination Fig. 5: Influence of high temperature annealing on the CVD ORNL polycrystalline 3 C-SiC. Fig. 6: Influence of high temperature annealing on the polycrystalline 3 C-SiC layer of PBMR TRISO CP batches D and E...

  2. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  3. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  4. Evaluation of design, leak monitoring, dnd NDEA strategies to assure PBMR Helium pressure boundary reliability - HTR2008-58037

    International Nuclear Information System (INIS)

    Fleming, K. N.; Smit, K.

    2008-01-01

    This paper discusses the reliability and integrity management (RIM) strategies that have been applied in the design of the PBMR passive metallic components for the helium pressure boundary (HPB) to meet reliability targets and to evaluate what combination of strategies are needed to meet the targets. The strategies considered include deterministic design strategies to reduce or eliminate the potential for specific damage mechanisms, use of an on-line leak monitoring system and associated design provisions that provide a high degree of leak detection reliability, and periodic nondestructive examinations combined with repair and replacement strategies to reduce the probability that degradation would lead to pipe ruptures. The PBMR RIM program for passive metallic piping components uses a leak-before-break philosophy. A Markov model developed for use in LWR risk-informed in-service inspection evaluations was applied to investigate the impact of alternative RIM strategies and plant age assumptions on the pipe rupture frequencies as a function of rupture size. Some key results of this investigation are presented in this paper. (authors)

  5. Pebble bed modular reactor - The first Generation IV reactor to be constructed

    International Nuclear Information System (INIS)

    Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

    2004-01-01

    Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

  6. RETOS DEL APRENDIZAJE BASADO EN PROBLEMAS

    OpenAIRE

    Carlos Antonio Poot-Delgado

    2013-01-01

    Se realiza una breve descripción de la viabilidad y retos de llevar a la práctica docente el aprendizaje basado en problemas, con una estructura organizada, donde se plasman ideas bien fundamentadas con claridad y precisión. Asimismo, se plantean reflexiones pertinentes, mediante ejercicios alusivos y ubicados en la práctica docente. Es decir, una forma de trabajo que puede ser usada por el edu- cador en una parte de su curso, combinado con otras técnicas didácticas y de- limitando los objeti...

  7. Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

    Science.gov (United States)

    Mkhabela, Peter Tshepo

    The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis

  8. Pebble bed modular reactors versus other generation technologies. Costs and challenges for South Africa

    International Nuclear Information System (INIS)

    Grubert, Emily; Parks, Brian; Schneider, Erich; Sekar, Srinivas

    2011-01-01

    South Africa is Africa's major economy, with plans to double its electricity generation capacity by 2026. South Africa has spent almost two decades developing a nuclear reactor known as a Pebble Bed Modular Reactor (PBMR), which could provide substantial benefits to the electricity grid but was recently mothballed due to high costs. This work estimates the lifecycle financial costs of South African PBMRs, then compares these costs to those of five other generation options: coal, nuclear as pressurized water reactors (PWRs), wind, and solar as photovoltaics (PV) or concentrating solar power (CSP). Each technology is evaluated with low, base case, and high assumptions for capital costs, construction time, and interest rates. Decommissioning costs, project lifetime, capacity factors, and sensitivity to carbon price are also considered. PBMR could be cost competitive with coal under certain low cost conditions, even without a carbon price. However, international lending practices and other factors suggest that a high capital cost, high interest rate nuclear plant is likely to be competing with a low capital cost, low interest rate coal plant in a market where cost recovery is challenging. PBMR could potentially become more competitive if low rate international loans were available to nuclear projects or became unavailable to coal projects. (author)

  9. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

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

  10. Membrane reactor for water detritiation: a parametric study on operating parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mascarade, J.; Liger, K.; Troulay, M.; Perrais, C. [CEA, DEN, DTN/STPA/LIPC, Centre de Cadarache, Saint-Paul-lez-Durance (France); Joulia, X.; Meyer, X.M. [Universite de Toulouse, INPT, UPS, Laboratoire de Genie Chimique, Toulouse (France); CNRS, Laboratoire de Genie Chimique, Toulouse (France)

    2015-03-15

    This paper presents the results of a parametric study done on a single stage finger-type packed-bed membrane reactor (PBMR) used for heavy water vapor de-deuteration. Parametric studies have been done on 3 operating parameters which are: the membrane temperature, the total feed flow rate and the feed composition through D{sub 2}O content variations. Thanks to mass spectrometer analysis of streams leaving the PBMR, speciation of deuterated species was achieved. Measurement of the amounts of each molecular component allowed the calculation of reaction quotient at the packed-bed outlet. While temperature variation mainly influences permeation efficiency, feed flow rate perturbation reveals dependence of conversion and permeation properties to contact time between catalyst and reacting mixture. The study shows that isotopic exchange reactions occurring on the catalyst particles surface are not thermodynamically balanced. Moreover, the variation of the heavy water content in the feed exhibits competition between permeation and conversion kinetics.

  11. 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)

  12. Risk-informed design of a pebble bed gas reactor

    International Nuclear Information System (INIS)

    Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

    2003-01-01

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

  13. OTTO-PAP: An alternative option to the PBMR fuelling philosophy

    International Nuclear Information System (INIS)

    Mulder, E.; Teuchert, E.

    1997-01-01

    Once Through Then Out, Power Adjusted by Poison (OTTO-PAP) fuelling of a high temperature pebble-bed reactor offers a simple alternative to the MEDUL (Mehrfachdurchlauf = German for multi-pass) fuelling regime followed in pebble bed reactor designs to date. The prerequisite for a modular reactor unit of maximum power output, subject to observing passive safety characteristics is a sufficiently flat axial neutron flux profile. This is achieved by introducing B 4 C coated particles of pre-calculated size and packing density within the fuel spheres. In accordance with AVR operating practise the temperature profile is radially equalised by introducing a 2-zone core loading. Adding pure graphite spheres loosely into the centre column area of the core effectively reduced the maximum power in the middle. Increasing the reactor diameter is enabled by the introduction of noses. A 3-D geometric modeller developed in cylindrical co-ordinates enables a given flow description of the pebbles adjacent to the nose boundaries and in the vicinity of the shut down/control rods. After translation of the geometric data the neutronic behaviour of the reactor is followed in 3-D by the CITATION code. This study is aimed towards achieving an optimal core layout with a LEU (Low Enriched Uranium) fuel cycle. Physical properties of the OTTO-PAP, 150 MWt reference design is reported, while computations performed observe results obtained by the reference HTR-MODUL design. (author)

  14. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

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

  15. Development of a system based in a digital signal processor (DSP) for a simulator of power regulation in a reactor: first stage; Desarrollo de un sistema basado en un DSP para un simulador de regulacion de potencia en un reactor: 1. etapa

    Energy Technology Data Exchange (ETDEWEB)

    Benitez R, J.S.; Perez C, B. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Municipio de Ocoyoacac, 52045 Estado de Mexico (Mexico)

    2002-07-01

    The first stage of the development of a digital system based on a DSP is presented which forms part of an hybrid simulator for the power regulation in am model of the punctual kinetics of a TRIGA reactor type. The DSP performs the regulation, using a Mandami type algorithm of diffuse control. In the algorithm, the universe of the output variable is discretized for performing in an unique stage the aggregation functions and dis-diffusization. (Author)

  16. Process heat cogeneration using a high temperature reactor

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Ramirez, Ramon; Valle, Edmundo del; Castillo, Rogelio

    2014-01-01

    Highlights: • HTR feasibility for process heat cogeneration is assessed. • A cogeneration coupling for HTR is proposed and process heat cost is evaluated. • A CCGT process heat cogeneration set up is also assessed. • Technical comparison between both sources of cogeneration is performed. • Economical competitiveness of the HTR for process heat cogeneration is analyzed. - Abstract: High temperature nuclear reactors offer the possibility to generate process heat that could be used in the oil industry, particularly in refineries for gasoline production. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product and if the cost of this subproduct will be competitive with other alternatives. The current study assesses the likeliness of generating process heat from Pebble Bed Modular Reactor to be used for a refinery showing different plant balances and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor where the cycle configuration to transport the heat of the reactor to the process plant plays an important role in the cycle efficiency and in the plant economics. The results of this study show that the PBMR would be most competitive when capital discount rates are low (5%), carbon prices are high (>30 US$/ton), and competing natural gas prices are at least 8 US$/mmBTU

  17. Process heat cogeneration using a high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Gustavo, E-mail: gustavoalonso3@gmail.com [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ramirez, Ramon [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Valle, Edmundo del [Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Castillo, Rogelio [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico)

    2014-12-15

    Highlights: • HTR feasibility for process heat cogeneration is assessed. • A cogeneration coupling for HTR is proposed and process heat cost is evaluated. • A CCGT process heat cogeneration set up is also assessed. • Technical comparison between both sources of cogeneration is performed. • Economical competitiveness of the HTR for process heat cogeneration is analyzed. - Abstract: High temperature nuclear reactors offer the possibility to generate process heat that could be used in the oil industry, particularly in refineries for gasoline production. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product and if the cost of this subproduct will be competitive with other alternatives. The current study assesses the likeliness of generating process heat from Pebble Bed Modular Reactor to be used for a refinery showing different plant balances and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor where the cycle configuration to transport the heat of the reactor to the process plant plays an important role in the cycle efficiency and in the plant economics. The results of this study show that the PBMR would be most competitive when capital discount rates are low (5%), carbon prices are high (>30 US$/ton), and competing natural gas prices are at least 8 US$/mmBTU.

  18. Office of Nuclear Regulatory Research summary of advanced reactors activities, June 4, 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Pre-application interactions with potential licensee applicants will help NRC prepare for future submittals, through the development of the infrastructure necessary for licensing application reviews. RES has the lead for non-LWR advanced reactor pre-application initiatives and longer-range new technology initiatives. An advanced reactor group has been formed in REAHFB, and is currently performing a pre-application review of Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre application interaction include General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) design. RES advanced reactors activities also include participation as an observer in DOE's Generation IV initiative. Pre-Application review objectives include the development of regulatory guidance, licensing approach, and technology-basis expectations for licensing advanced designs, including identifying significant technology, design, safety, licensing and policy issues that would need to be addressed in the licensing process. The presentation described the pre-application process for the Exelon PBMR. NRC first identifies additional information following topical meetings with Exelon, and Exelon formally documents and submits required topical Information. The staff then develops a preliminary assessment and drafts a response which is followed by stakeholder input and comments at a public workshop. Preliminary assessments are discussed with ACRS and ACNW, and Commission papers are written which provide staff positions and recommendations on proposed policy decisions. Some of the significant areas for the PBMR include: Process Issues, Legal and Financial Issues; Regulatory Framework; Fuel Performance and Qualification; Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials); Fuel Cycle Safety Areas; PRA, SSC Safety Classification; PBMR Prototype Testing

  19. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning

  20. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

  1. Study for Safeguards Challenges to the Most Probably First Indonesian Future Power Plant of the Pebble Bed Modular Reactor

    International Nuclear Information System (INIS)

    Susilowati, E.

    2015-01-01

    In the near future Indonesia, the fourth most populous country, plans to build a small size power plant most probably a Pebble Bed Modular Reactor PBMR. This first nuclear power plant (NPP) is aimed to provide clear picture to the society in regard to performance and safety of nuclear power plant operation. Selection to the PBMR based on several factor including the combination of small size of the reactor and type of fuel allowing the use of passive safety systems, resulting in essential advantages in nuclear plant design and less dependence on plant operators for safety. In the light of safeguards perspective this typical reactor is also quite difference with previous light water reactor (LWR) design. From the fact that there are a small size large number of elements present in the reactor produced without individual serial numbers combine to on-line refueling same as the CANDU reactor, enforcing a new challenge to safeguards approach for this typical reactor. This paper discusses a bunch of safeguards measures have to be prepared by facility operator to support successfully international nuclear material and facility verification including elements of design relevant to safeguards need to be accomplished in consultation to the regulatory body, supplier or designer and the Agency/IAEA such as nuclear material balance area and key measurement point; possible diversion scenarios and safeguards strategy; and design features relevant to the IAEA equipment have to be installed at the reactor facility. It is deemed that result of discussion will alleviate and support the Agency approaching safeguards measure that may be applied to the purpose Indonesian first power plant of PBMR construction and operation. (author)

  2. 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)

  3. New advanced small and medium nuclear power reactors: possible nuclear power plants for Australia

    International Nuclear Information System (INIS)

    Dussol, R.J.

    2003-01-01

    In recent years interest has increased in small and medium sized nuclear power reactors for generating electricity and process heat. This interest has been driven by a desire to reduce capital costs, construction times and interest during construction, service remote sites and ease integration into small grids. The IAEA has recommended that the term 'small' be applied to reactors with a net electrical output less than 300 MWe and the term 'medium' to 300-700 MWe. A large amount of experience has been gained over 50 years in the design, construction and operation of small and medium nuclear power reactors. Historically, 100% of commercial reactors were in these categories in 1951-1960, reducing to 21% in 1991-2000. The technologies involved include pressurised water reactors, boiling water reactors, high temperature gas-cooled reactors, liquid metal reactors and molten salt reactors. Details will be provided of two of the most promising new designs, the South African Pebble Bed Modular Reactor (PBMR) of about 110 MWe, and the IRIS (International Reactor Innovative and Secure) reactor of about 335 MWe. Their construction costs are estimated to be about US$l,000/kWe with a generating cost for the PBMR of about US1.6c/kWh. These costs are lower than estimated for the latest designs of large reactors such as the European Pressurised Reactor (EPR) designed for 1,600 MWe for use in Europe in the next decade. It is concluded that a small or medium nuclear power reactor system built in modules to follow an increasing demand could be attractive for generating low cost electricity in many Australian states and reduce problems arising from air pollution and greenhouse gas emissions from burning fossil fuels

  4. APRENDIZAJE BASADO EN PROBLEMAS Y RAZONAMIENTO BASADO EN CASOS EN LA ENSEÑANZA

    Directory of Open Access Journals (Sweden)

    Juan Pedro Febles Rodríguez

    2002-05-01

    Full Text Available

    El el trabajo se hace una valoración sobre la utilización del Razonamiento Basado en Caso, como eficaz complemento en la formación profesional del alumno de Medicina, ya que este tipo de sistema utiliza un mecanismo de razonamiento por analogías o asociaciones de forma automática, muy similar a como lo realiza el humano. Esto permitirá que ante la presentación de un problema (teniendo en cuenta los principales aspectos destacados sobre el Aprendizaje Basado en Problemas, PBL, en las distintas bases de conocimientos que se conformen y que estarán constituidas por casos reales o supuestos, prototipos y excepcionales, en número suficiente y aportadas por los profesores, quienes fungirán como expertos, los educandos busquen respuestas a sus inquietudes.

  5. A CFD Study on Inlet Plenum Flow Field of Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Lee, Won Jae; Chang, Jong Hwa

    2005-01-01

    High temperature gas cooled reactor, largely divided into two types of PBR (Pebble Bed Reactor) and PMR (Prismatic Modular Reactor), has becomes great interest of researchers in connection with the hydrogen production. KAERI has started a project to develop the gas cooled reactor for the hydrogen production and has been doing in-depth study for selecting the reactor type between PBR and PMR. As a part of the study, PBMR (Pebble Bed Modular Reactor) was selected as a reference PBR reactor for the CFD analysis and the flow field of its inlet plenum was simulated with computational fluid dynamics program CFX5. Due to asymmetrical arrangement of pipes to the inlet plenum, non-uniform flow distribution has been expected to occur, giving rise to non-uniform power distribution at the core. Flow fields of different arrangement of inlet pipes were also investigated, as one of measures to reduce the non-uniformity

  6. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Bjornard, Trond; Hockert, John

    2011-01-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC and A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC and A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC and A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR (Pty) and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC and A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR and D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present

  7. Influence of temperature on the micro-and nano-structures of experimental PBMR TRISO coated particles: A comparative study - HTR2008-58189

    International Nuclear Information System (INIS)

    Van Rooyen, I. J.; Neethling, J. H.; Mahlangu, J.

    2008-01-01

    The PBMR fuel consists of TRISO Coated Particles (CPs) in a graphite matrix. The three layer system, IPyC-SIC-OPyC, forms the primary barrier to fission product release, with the SiC layer acting as the main pressure boundary of the particle. The containment of fission products inside the CPs is however a function of the operating temperature and microstructure of the SiC layer. During accident conditions, the CPs will reach higher temperatures than normal operating conditions. The Fuel Design Dept. of PBMR is therefore investigating various characteristics of the SiC layer, especially nano characteristics at variant conditions. The integrity of the interface between the SiC and Inner PyC layers is also important for fission product retention and therefore interesting TEM images of this region of the experimental PBMR TRISO particles will be shown. Transmission electron microscope (TEM) images of the microstructure of TRISO coated particles of three different experimental batches after annealing will be discussed. Particles annealed at 1980 deg. C for 1 hour revealed that the inner PyC layer de-bonded from the SiC layer. Changes observed in the diffraction rings are evidence that the PyC structure is becoming organized or anisotropic. The SiC layer, on the other hand, did not show any changes as a result of annealing. Only the cubic 3C-SiC phase was observed for a limited number of grains analyzed. The nano hardness and elasticity measurements of the three test batches were done using a CSM Nano Hardness Tester. These results are compared to indicate possible differences between the 1 hour and 5 hour annealing time as a function of annealing temperature from 1000 deg. C to 1980 deg. C. The variation of hardness and elasticity as a function of temperature for the three experimental batches are identified and discussed. This preliminary TEM investigation and nano hardness measurements have contributed new knowledge about the effect of high temperature annealing on

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  9. Sistema de inferencia difusa basado en relaciones Booleanas

    Directory of Open Access Journals (Sweden)

    Helbert Eduardo Espitia Cuchango

    2010-09-01

    Full Text Available Este documento describe la estructura de un sistema de inferencia difusa basado en relaciones booleanas. La teoría relacionada con lógica y conjuntos booleanos es una buena herramienta para el diseño de automatismos y sistemas digitales. Una variación con la cual se busca mejorar los sistemas basados en automatismos consiste en emplear conjuntos difusos en lugar de booleanos. Lo anterior se realiza con el objetivo de tener una acción continua en el actuador del automatismo. Al realizar esta variación y al aplicar la metodología de diseño de los sistemas de automatismos, aparecen los sistemas de inferencia difusa basados en relaciones booleanas.Aunque inicialmente esta propuesta se realizó considerando sistemas de automatismos, se observa que es posible extenderla a sistemas de inferencia difusa.

  10. Posicionamiento de robots basado en visión

    OpenAIRE

    Marcano Gamero, Cosme Rafael

    2007-01-01

    Se ha hecho una revisión somera de algunos métodos de localización de robots basados en la visión computarizada. Estos métodos incluyen: a. uso de marcas de referencias fijadas a tierra b. modelos de objetos c. mapas d. construcción de mapas basado en las características observadas. Todos estos métodos están en experimentación. Se han hecho significativos avances en la fusión de varias de estas técnicas como aquellos basados en la odometría, en conjunto con técnicas de construcción de mapas b...

  11. Radioactive waste management plan for the PBMR (Pty) Ltd fuel plant

    International Nuclear Information System (INIS)

    Makgae, Mosidi E.

    2009-01-01

    The Pebble Bed Modular Reactor (Pty) Ltd Fuel Plant (PFP) radioactive waste management plan caters for waste from generation, processing through storage and possible disposal. Generally, the amount of waste that will be generated from the PFP is Low and Intermediate Level Waste. The waste management plan outlines all waste streams and the management options for each stream. It also discusses how the Plant has been designed to ensure radioactive waste minimisation through recycling, recovery, reuse, treatment before considering disposal. Compliance to the proposed plan will ensure compliance with national legislative requirements and international good practice. The national and the overall waste management objective is to ensure that all PFP wastes are managed appropriately by utilising processes that minimize, reduce, recover and recycle without exposing employees, the public and the environment to unacceptable impacts. Both International Atomic Energy Agency (IAEA) and Department of Minerals and Energy (DME) principles act as a guide in the development of the strategy in order to ensure international best practice, legal compliance and ensuring that the impact of waste on employees, environment and the public is as low as reasonably achievable. The radioactive waste classification system stipulated in the Radioactive Waste Management Policy and Strategy 2005 will play an important role in classifying radioactive waste and ensuring that effective management is implemented for all waste streams, for example gaseous, liquid or solid wastes.

  12. Turbulence-induced heat transfer in PBMR core using LES and RANS

    International Nuclear Information System (INIS)

    Lee, Jung-Jae; Yoon, Su-Jong; Park, Goon-Cherl; Lee, Won-Jae

    2007-01-01

    This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution. (author)

  13. Cogeneration using a nuclear reactor to generate process heat

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Ramirez, Ramon

    2009-01-01

    Some of the new nuclear reactor technologies (Generation III+) are claiming the production of process heat as an additional value to electricity generation. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product. The current study assess the likeliness of generate process heat from a Pebble Bed Modular Reactor to be used for a refinery showing different plant balance and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor and also the challenges that this option has. (author)

  14. Simulador electromagnético basado en FDTD

    OpenAIRE

    Inclan Alonso, Jose Manuel

    2012-01-01

    El presenta trabajo fin de master tiene como objetivo el desarrollo de un simulador electromagnético basado en FDTD. El simulador incluye fronteras abiertas basadas en PML y transformada de campo cercano a lejano en el dominio del tiempo.

  15. Control predictivo basado en predictores borrosos

    Directory of Open Access Journals (Sweden)

    MIGUEL ANGEL RODRIGUEZ BORROTO

    2007-01-01

    Full Text Available En el presente artículo se expone un algoritmo para aplicar las técnicas de control predictivo lineal basadas en el modelo (MPC al caso de procesos no lineales utilizando un predictor borroso (fuzzy, en base a la estructura Takagi_Sugeno_Kang_dinámica y bajo el principio de utilización de modelos lineales locales por tramos. Ello conduce inherentemente, a un proceso de adaptación de la matriz dinámica del sistema en cada periodo de muestreo, lo cual se considera novedoso en relación al MPC clásico. Se exponen los resultados de su aplicación al caso de un control de posición que utiliza un servo-motor de corriente continua con zona muerta y a un reactor continuo de tanque con agitador con reacción exotérmica. Los resultados son satisfactorios

  16. Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

    International Nuclear Information System (INIS)

    In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

    2008-01-01

    Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

  17. MAPA CONCEPTUAL PARA EL APRENDIZAJE BASADO EN PROBLEMAS

    Directory of Open Access Journals (Sweden)

    YURI GORBANEFF

    2009-01-01

    Full Text Available Se presentan los resultados del experimento sobre el uso de mapas conceptuales en el contexto del aprendizaje basado en problemas (ABP. El experimento fue realizado en la Universidad Javeriana (Bogotá, Colombia en el pregrado de administración. Se encontró que el mapa conceptual mejora el aprendizaje pero no afecta la percepción de los alumnos del método de ABP. Se recomienda incluir mapas conceptuales en los ejercicios de ABP.

  18. Mapa Conceptual Para El Aprendizaje Basado En Problemas

    Directory of Open Access Journals (Sweden)

    Yuri Gorbaneff

    2009-01-01

    Full Text Available Se presentan los resultados del experimento sobre el uso de mapas conceptuales en el contexto del aprendizaje basado en problemas (ABP. El experimento fue realizado en la Universidad Javeriana (Bogotá, Colombia en el pregrado de administración. Se encontró que el mapa conceptual mejora el aprendizaje pero no afecta la percepción de los alumnos del método de ABP. Se recomienda incluir mapas conceptuales en los ejercicios de ABP.

  19. Sensores electroquímicos basados en nanomateriales de carbono

    OpenAIRE

    Remesal García, Lucía

    2017-01-01

    Caracterización de tres sensores basados en materiales de nanocarbono mediante el análisis de diferentes compuestos. El objetivo del proyecto ha sido analizar que sensor es el más sensible para detectar los compuestos electroactivos en soluciones. Se han utilizado tres tipos diferentes de sensores: Electrodo de carbono modificado con nanotubos de carbono (CNT), Electrodo de carbono modificado con Nanofibras de carbono grafitizadas (CNF) y Electrodo de carbono modificado con grafeno (GPH). El ...

  20. South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

    International Nuclear Information System (INIS)

    Joubert, J.; Kohtz, N.; Coe, I.

    2008-01-01

    It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /l/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient

  1. Characteristic behaviour of Pebble Bed High Temperature Gas-cooled Reactors during water ingress events

    International Nuclear Information System (INIS)

    Khoza, Samukelisiwe N.; Serfontein, Dawid E.; Reitsma, Frederik

    2014-01-01

    The presence of water on the tube-side of the steam generators in high temperature gas-cooled reactors (HTGRs) with indirect cycle layouts presents a possibility for a penetration of neutron moderating steam into the core, which may cause a power excursion. This article presents results on the effect of water ingress into the core of the two South African Pebble Bed Modular Reactor design concepts, i.e. the PBMR-200 MW th and the PBMR-400 MW th developed by PBMR SOC Ltd. The VSOP 99/05 suite of codes was used for the simulation of this event. Partial steam vapour pressures were added in stages into the primary circuit in order to investigate the effect of water ingress on reactivity, power profiles and thermal neutron flux profiles. The effects of water ingress into the core are explained by increased neutron moderation, due to the addition of 1 H, which leads to a decrease in resonance capture by 238 U and therefore an increase in the multiplication factor. The more effective moderation of neutrons by definition reduces the fast neutron flux and increases the thermal flux in the core, i.e. leads to a softer spectrum. The more effective moderation also increases the average increase in lethargy between collisions of a neutron with successive fuel kernels, which reduces the probability for neutron capture in the radiative capture resonances of 238 U. The resulting higher resonance escape probability also increases the thermal flux in the core. The softening of the neutron spectrum leads to an increased effective microscopic fission cross section in the fissile isotopes and thus to increased neutron absorption for fission, which reduces the remaining number of neutrons that can diffuse into the reflectors. Therefore water ingress into the core leads to a reduced thermal neutron flux in the reflectors. The power density spatial distribution behaved similarly to the thermal neutron flux in the core. Analysis of possible mechanisms was conducted. The results show that

  2. Teoría redox mediante aprendizaje basado en problemas

    OpenAIRE

    Pérez Lemus, Nereida

    2016-01-01

    La ciencia (en nuestro caso, la Química) tiene una valoración positiva por parte de la sociedad pero los estudiantes muestran poco interés por las materias de carácter científico. Ante esta situación, la metodología tradicional (clase magistral) está siendo sustituida por otros métodos, entre ellos, el aprendizaje basado en problemas, para conseguir aumentar el interés y la motivación de los alumnos por estas materias. El objetivo consiste en que sean capaces de alcanzar las competencias clav...

  3. Sistemas de Inteligencia Web basados en redes sociales

    OpenAIRE

    Rosa Troyano, Fco. Fernando de la; Martínez Gasca, Rafael

    2007-01-01

    El Análisis de las Redes Sociales (ARS) es un área que está emergiendo como imprescindible en los procesos de toma de decisiones. Su capacidad para analizar e intervenir una red social puede ser aprovechada para implantar tareas de vigilancia en los sistemas de inteligencia de un centro de investigación o una empresa de base tecnológica. El objetivo de este trabajo es realizar una propuesta para diseñar sistemas de inteligencia web basados en redes sociales. El primer obstáculo para implantar...

  4. Sistemas de Inteligencia Web basados en Redes Sociales

    OpenAIRE

    de la Rosa Troyano, Fco. Fernando; Martínez Gasca, Rafael

    2007-01-01

    El Análisis de las Redes Sociales (ARS) es un área que está emergiendo como imprescindible en los procesos de toma de decisiones. Su capacidad para analizar e intervenir una red social puede ser aprovechada para implantar tareas de vigilancia en los sistemas de inteligencia de un centro de investigación o una empresa de base tecnológica. El objetivo de este trabajo es realizar una propuesta para diseñar sistemas de inteligencia web basados en redes sociales. El primer obstáculo para implantar...

  5. PROPIEDADES ELECTRONICAS DE SISTEMAS BASADOS EN GRAFENO BICAPA

    OpenAIRE

    SUAREZ MORELL; ERIC; SUAREZ MORELL; ERIC

    2011-01-01

    La estructura electrónica de una monocapa de Carbono fue estudiada inicialmente[l ] como una primera aproximación al estudio de las propiedades del Grafito, sin embargo no fue sino hasta el 2004 que fue sintetizado experimentalmente[2]. La monocapa de Carbono ha recibido el nombre en la literatura de Grafeno palabra proveniente del Grafito y del Fullereno, otro de los alotropos del Carbono. A partir del 2004 el estudio de sistemas basados en Grafeno se ha convertido en uno de l...

  6. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  7. Itinerarios de aprendizaje flexibles basados en mapas conceptuales

    Directory of Open Access Journals (Sweden)

    Olga Lucía Agudelo

    2015-07-01

    Full Text Available Se estudia el uso de itinerarios de aprendizaje  basados en mapas conceptuales como una propuesta para un diseño instruccional más flexible que potencie el aprendizaje y se centre en el estudiante, generando procesos no lineales, caracterizando sus elementos, estableciendo relaciones entre ellos y configurando un modelo general con especificaciones para  cada nivel de educación. Mediante una metodología construida sobre el modelo SAM (Successive Aproximation Model, se ha estudiado el proceso de diseño, implementación y evaluación de itinerarios de aprendizaje, en los que se representan en mapas conceptuales y de manera organizada y no lineal, conjuntos de actividades  que permiten el desarrollo de las competencias que deben comprenderse, dominarse y demostrarse. Los resultados obtenidos muestran la adecuación del itinerario de aprendizaje basado en mapas conceptuales a las características de los sujetos, resolviendo situaciones de la realidad mediante la construcción y creación de nuevos esquemas y formas de gestión del conocimiento, al mismo tiempo que aporta reflexión sobre los principios del diseño curricular utilizando TIC y genera autonomía en el proceso de aprendizaje.

  8. Phenomenological modeling and study of a catalytic membrane reactor for water detritiation

    International Nuclear Information System (INIS)

    Mascarade, Jeremy

    2015-01-01

    Tritium is produced in light and heavy water reactor fuel by ternary fission or neutron activation. This by-product is used as fuel in fusion fuel reactors such as JET in Culham or ITER in Cadarache (France). The growing interest of this research area will make the tritium fluxes increase; it is then worth addressing the question of its future whether it will be used or flushed out from liquid and gaseous effluents or waste. This thesis studies the recovery of tritium as fuel for fusion machines by means of packed bed membrane reactor (PBMR). Such a reactor combines catalytic conversion of tritiated water thanks to isotope exchange with hydrogen according to the reversible reaction Q 2 O+H 2 ↔H 2 O+Q 2 (Q=H,D or T) and selective permeation of Q 2 through Pd-based membrane. In fact, palladium has the ability to bond with hydrogen isotopes, creating a selective permeation barrier. In the PBMR, thanks to the reaction products withdrawal, these permeation fluxes drive the heavy water conversion rate, to higher values than those reached in conventional fixed bed reactors (Le Chatelier's law). In order to study PBMRs, the CEA has built a test bench, using deuterium instead of tritium, allowing the analysis of their conversion and separation performances at the laboratory scale. An in-house method has been developed to determine simultaneously hydrogen and water isotopologues content by mass spectrometer analysis. It was experimentally shown that the activity of Ni-based catalyst used in this study was sufficient to allow the isotope exchange reactions to reach their thermodynamic equilibrium in a very short time. In addition, hydrogen permeation flux was shown to follow a Richardson's law. Sensitivity studies performed on the PBMR's main operating parameters revealed that its global performance (i.e. de-deuteration factor) increases with the temperature, the transmembrane pressure difference, the sweep gas flow rate and the residence time in the catalyst

  9. Current status and results of the PBMR -Pebble Box- benchmark within the framework of the IAEA CRP5 - 341

    International Nuclear Information System (INIS)

    Reitsma, F.; Tyobeka, B.

    2010-01-01

    The verification and validation of computer codes used in the analysis of high temperature gas cooled pebble bed reactor systems has not been an easy goal to achieve. A limited amount of tests and operating reactor measurements are available. Code-to-code comparisons for realistic pebble bed reactor designs often exhibit differences that are difficult to explain and are often blamed on the complexity of the core models or the variety of analysis methods and cross section data sets employed. For this reason, within the framework of the IAEA CRP5, the 'Pebble Box' benchmark was formulated as a simple way to compare various treatments of neutronics phenomena. The problem is comprised of six test cases which were designed to investigate the treatments and effects of leakage and heterogeneity. This paper presents the preliminary results of the benchmark exercise as received during the CRP and suggests possible future steps towards the resolution of discrepancies between the results. Although few participants took part in the benchmarking exercise, the results presented here show that there is still a need for further evaluation and in-depth understanding in order to build the confidence that all the different methods, codes and cross-section data sets have the capability to handle the various neutronics effects for such systems. (authors)

  10. Presentation summary, safety design aspects and U.S. licensing challenges of the pebble bed modular reactor

    International Nuclear Information System (INIS)

    Sproat, Ward; Slabber, Johan

    2001-01-01

    This presentation consists of three sections: An overview of the status of the PBMR project in South Africa, a review of the design features and philosophy being utilized to design the PBMR, and a summary of the key licensing issues that Exelon has identified in assessing the licensability of the PBMR for application in this country

  11. Sensores de fibra óptica basados en resonancias electromagnéticas

    OpenAIRE

    López Lambás, Sergio

    2011-01-01

    Este trabajo surge como continuación a los trabajos basados en LMR ya realizados con el objetivo de profundizar en el estudio y fabricación de sensores de fibra óptica basados en resonancias electromagnéticas utilizando diferentes materiales y técnicas. En concreto el proyecto perseguirá los siguientes objetivos: Desarrollar sensores de fibra óptica basados en resonancias electromagnéticas con recubrimientos metálicos y su aplicación a la detección de campo magnético. Estudiar y mejorar la fa...

  12. Metaphysics methods development for high temperature gas cooled reactor analysis

    International Nuclear Information System (INIS)

    Seker, V.; Downar, T. J.

    2007-01-01

    Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

  13. hipermedia adaptativos educativos basados en estilos de aprendizaje

    Directory of Open Access Journals (Sweden)

    Helmut Leighton Álvarez

    2005-01-01

    Full Text Available Este trabajo es una propuesta metodológica para determinar atributos y métricas de calidad en sistemas hipermedia adaptativos educativos basados en estilos de aprendizaje, específicamente en los estilos activo, reflexivo, teórico y pragmático. Estos atributos y métricas están referidos únicamente a la interacción del usuario-estudiante con el sistema, es decir, desde la óptica puramente educativa y no desde el punto de vista de la herramienta informática como tal. Para ello comienza su análisis desde las características del estilo de aprendizaje y a partir de ellas se procede al establecimiento de los atributos de acuerdo a las estrategias instruccionales que le correspondan. Finalmente, se definen las métricas necesarias para cada uno de los atributos, estableciendo el o los tipos de variables que involucran, sus unidades de medida y escalas. De esta manera se desarrolla una metodología de cascada de determinación de atributos y métricas

  14. Aprendizaje basado en problemas aplicado a las lenguas de especialidad

    Directory of Open Access Journals (Sweden)

    Mª Ángeles Andreu-Andrés

    2010-04-01

    Full Text Available El ámbito universitario es un campo propicio para implementar el aprendizaje activo y el trabajo en equipo de los estudiantes de ingeniería, de modo que su práctica fomente la adquisición de destrezas y competencias que les sean útiles académica y profesionalmente. Una de estas técnicas de aprendizaje activo es el “Aprendizaje Basado en Problemas” (ABP. Si bien su uso se centra mayoritariamente en asignaturas de ciencias, este trabajo presenta parte de los resultados de una experiencia llevada a cabo dentro de una asignatura de inglés para fines académicos y profesionales. En ella, el ABP, el trabajo en equipo y la lengua de especialidad se entremezclan de manera natural con el pensamiento crítico, las presentaciones orales y la evaluación de la participación de cada miembro de los equipos a lo largo de las tareas que lo conforman. Los resultados que aquí se presentan ofrecen el producto diseñado y consensuado por los once grupos participantes para evaluar una presentación oral así como para autoevaluar y evaluar la participación de cada miembro de los equipos durante el proceso.

  15. APRENDIZAJE BASADO EN PROBLEMAS (ABP: UNA EXPERIENCIA PEDAGOGICA EN MEDICINA

    Directory of Open Access Journals (Sweden)

    Claudio Lermanda S.

    2007-01-01

    Full Text Available La enseñanza tradicional de la Medicina ha sido cuestionada por el avance tecnológico y el exponencial aumento de información médica accesible, obligando a reformular los curricula médicos bajo la mirada de las reformas educacionales que a escala global vienen desarrollándose desde el Proceso de Bolonia. Una propuesta curricular nueva señalada por Harden (1984 y conocida como modelo SPICES permite incorporar una metodología didáctica más apropiada para el proceso de enseñanza - aprendizaje formativo e integrador que demanda la sociedad actual. En este nuevo escenario, muchas Facultades de Medicina han incorporado el Aprendizaje Basado en Problemas (ABP como herramienta de aproximación al modelo señalado, por su carácter integrador de conocimientos, destrezas y actitudes que facilitan la adquisición de las competencias clínicas necesarias para el ejercicio profesional futuro. Este trabajo resume las consideraciones y sugerencias relativas al modelo y a la metodología tras una experiencia de dos años de ABP en la Facultad de Medicina de la Universidad Católica de la Santísima Concepción.

  16. Validation of CATHARE for gas-cooled reactors

    International Nuclear Information System (INIS)

    Fabrice Bentivoglio; Ola Widlund; Manuel Saez

    2005-01-01

    Full text of publication follows: Extensively validated and qualified for light-water reactor safety studies, the thermo-hydraulics code CATHARE has been adapted to deal also with gas-cooled reactor applications. In order to validate the code for these novel applications, CEA (Commissariat a l'Energie Atomique) has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE is being validated against existing experimental data, in particular from the German power plant Oberhausen II and the South African Pebble-Bed Micro Model (PBMM). Oberhausen II, operated by the German utility EVO, is a 50 MW(e) direct-cycle Helium turbine plant. The power source is a gas burner rather than a nuclear reactor core, but the power conversion system resembles those of the GFR (Gas-cooled Fast Reactor) and other high-temperature reactor concepts. Oberhausen II was operated for more than 100 000 hours between 1974 and 1988. Design specifications, drawings and experimental data have been obtained through the European HTR project, offering a unique opportunity to validate CATHARE on a large-scale Brayton cycle. Available measurements of temperatures, pressures and mass flows throughout the circuit have allowed a very comprehensive thermohydraulic description of the plant, in steady-state conditions as well as during transients. The Pebble-Bed Micro Model (PBMM) is a small-scale model conceived to demonstrate the operability and control strategies of the South African PBMR concept. The model uses Nitrogen instead of Helium, and an electrical heater with a maximum rating of 420 kW. As the full-scale PBMR, the PBMM loop features three turbines and two compressors on the primary circuit, located on three separate shafts. The generator, however, is modelled by a third compressor on a separate circuit, with a

  17. DISEÑO DE SOFTWARE EDUCATIVO BASADO EN COMPETENCIAS

    Directory of Open Access Journals (Sweden)

    Manuel Fernando Caro Piñeres

    2009-01-01

    Full Text Available El presente artículo describe un modelo de diseño de software educativo basado en competencias, el cual presenta una visión integral del desarrollo de estas aplicaciones mediante la combinación de componentes pedagógicos, didácticos, multimediales y de ingeniería de software. El modelo sugerido se compone de cinco fases que detallan paso a paso los aspectos que se deben tener en cuenta para la creación de softwareeducativo. La fase inicial constituye la descripción del diseño educativo, en la cual se analiza la necesidad educativa, se plantean los objetivos de aprendizaje y se describen las competencias que se pretenden desarrollar con la aplicación; del diseño de éstas resultan las siguientes subfases: diseño de contenidos, diseño pedagógico y diseño de aprendizaje. Las competencias son el aspecto fundamental que abarca el desarrollo de este modelo, las cuales son primordiales para la realización de las fases a seguir conformadas por el diseño computacional y el diseño multimedial, estas se encargan del análisis y modelado del software, y del sistema de comunicación hombre-máquina. En la fase de producción se ensamblan los componentes elaborados o recolectados, según el caso. La última fase es la de aplicación, donde se hacen las pruebas de rigor para evaluar el desempeño del software en los contextos para los que fue desarrollado.

  18. Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

    International Nuclear Information System (INIS)

    Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

    2009-01-01

    Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

  19. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  20. Development of computational methods for the safety assessment of gas-cooled high-temperature and supercritical light-water reactors. Final report; Rechenmethoden zur Bewertung der Sicherheit von gasgekuehlten Hochtemperaturreaktoren und superkritischen Leichtwasserreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, S.; Cron, D. von der; Hristov, H.; Lerchl, G.; Papukchiev, A.; Seubert, A.; Sureda, A.; Weis, J.; Weyermann, F.

    2012-12-15

    This report documents developments and results in the frame of the project RS1191 ''Development of computational methods for the safety assessment of gas-cooled high temperature and supercritical light-water reactors''. The report is structured according to the five work packages: 1. Reactor physics modeling of gas-cooled high temperature reactors; 2. Coupling of reactor physics and 3-D thermal hydraulics for the core barrel; 3. Extension of ATHLET models for application to supercritical reactors (HPLWR); 4. Further development of ATHLET for application to HTR; 5. Further development and validation of ANSYS CFX for application to alternative reactor concepts. Chapter 4 describes the extensions made in TORT-TD related to the simulation of pebble-bed HTR, e.g. spectral zone buckling, Iodine-Xenon dynamics, nuclear decay heat calculation and extension of the cross section interpolation algorithms to higher dimensions. For fast running scoping calculations, a time-dependent 3-D diffusion solver has been implemented in TORT-TD. For the PBMR-268 and PBMR-400 as well as for the HTR-10 reactor, appropriate TORT-TD models have been developed. Few-group nuclear cross sections have been generated using the spectral codes MICROX- 2 and DRAGON4. For verification and validation of nuclear cross sections and deterministic reactor models, MCNP models of reactor core and control rod of the HTR-10 have been developed. Comparisons with experimental data have been performed for the HTR-10 first criticality and control rod worth. The development of the coupled 3-D neutron kinetics and thermal hydraulics code system TORT-TD/ATTICA3D is documented in chapter 5. Similar to the couplings with ATHLET and COBRA-TF, the ''internal'' coupling approach has been implemented. Regarding the review of experiments and benchmarks relevant to HTR for validation of the coupled code system, the PBMR-400 benchmarks and the HTR-10 test reactor have been selected

  1. CALIDAD DE LOS PROYECTOS DE SOFTWARE: REVISIONES UTILIZANDO RAZONAMIENTO BASADO EN CASOS

    Directory of Open Access Journals (Sweden)

    Martha Delgado Dapena

    2003-09-01

    Full Text Available

    En este trabajo se expone un sistema que permite planificar, controlar y dar seguimiento a las inspecciones realizadas a los proyectos de software, que utiliza el razonamiento basado en casos para planificar las inspecciones. Además, se presentan los conceptos fundamentales relacionados con las inspecciones dentro del sistema de aseguramiento de calidad, así como el procedimiento para llevarlas a cabo. También se expone una breve panorámica del razonamiento basado en casos.

  2. A systems CFD model of a packed bed high temperature gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Du Toit, C.G.; Rousseau, P.G.; Greyvenstein, G.P.; Landman, W.A.

    2006-01-01

    The theoretical basis and conceptual formulation of a comprehensive reactor model to simulate the thermal-fluid phenomena of the PBMR reactor core and core structures is given. Through a rigorous analysis the fundamental equations are recast in a form that is suitable for incorporation in a systems CFD code. The formulation of the equations results in a collection of one-dimensional elements (models) that can be used to construct a comprehensive multi-dimensional network model of the reactor. The elements account for the pressure drop through the reactor; the convective heat transport by the gas; the convection heat transfer between the gas and the solids; the radiative, contact and convection heat transfer between the pebbles and the heat conduction in the pebbles. Results from the numerical model are compared with that of experiments conducted on the SANA facility covering a range of temperatures as well as two different fluids and different heating configurations. The good comparison obtained between the simulated and measured results show that the systems CFD approach sufficiently accounts for all of the important phenomena encountered in the quasi-steady natural convection driven flows that will prevail after critical events in a reactor. The fact that the computer simulation time for all of the simulations was less than three seconds on a standard notebook computer also indicates that the new model indeed achieves a fine balance between accuracy and simplicity. The new model can therefore be used with confidence and still allow quick integrated plant simulations. (authors)

  3. Efforts of development on the next generation nuclear reactor in the Mitsubishi Heavy Industries, Ltd

    International Nuclear Information System (INIS)

    Mukai, Hiroshi

    2002-01-01

    At present, the Mitsubishi Heavy Industry, Ltd. (MHI) enters to development on APWR+ for a large-scale reactor, AP1000 and pebble bed modular reactor (PBMR) for middle- and small-scale one, and innovative one, under cooperation of power industries, manufacturers and institutes in and out of Japan. On APWR+, MHI occupies the most advanced position of conventional large-scale route, intends to carry out further upgrading of large capacity on a base of already developed 1500 MWe class APWR, and aims at further upgrading of economical efficiency. On the other reactor, as it becomes possible to perform value addition specific to the small-scale reactor with smaller output, it is planned to overcome its scale demerit by introducing more innovative techniques. And, on AP1000, it is intended to remove dynamic safety system by introducing a static one, to upgrade simplification of apparatus and reliability of safety system and to reduce its human factors. In addition, here was described on the next generation nuclear reactors under development. (G.K.)

  4. Sistema de detección de fallos basado en PC en calderas pirotubulares

    OpenAIRE

    Rivas Pérez, R.; Feliu Batlle, V.; Sotomayor Moriano, J.

    2005-01-01

    Se ofrece un sistema basado en PC para la detección de fallos en calderas pirotubulares. Se presentan los algoritmos que posibilitan la detección rápida de fallos abruptos en esta clase de plantas, los cuales se basan en la detección de cambios en los

  5. Aprendizaje basado en problemas y educación en ingeniería

    DEFF Research Database (Denmark)

    Numerosos reportes científicos alrededor del mundo identifican al aprendizaje basado en problemas (PBL) como una práctica educativa efectiva para que los estudiantes desarrollen habilidades y competencias para el ejercicio profesional durante su aprendizaje. En la comunidad académica se han formu...

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  7. Development of a tool for comparing different nuclear power reactor technologies: the Mexican choice

    International Nuclear Information System (INIS)

    Martin-del-Campo, C.; Francois, J.L.; Reyes, R.

    2007-01-01

    This paper describes a methodology which has allowed us to make a comparative assessment of nuclear power reactor options. The methodology was divided in 3 steps. The first step consists in searching of common indicators to be compared. A total of twenty indicators were considered and grouped in 3 main criteria. The second step is to obtain the values of all the indicators for each of the reactor technologies being compared. The third step is to utilize an aggregation method to integrate all the indicators in an overall qualification. Fuzzy Logic was selected as multi criteria aggregation method because it copes with imprecisely defined data; it can model non-linear functions of arbitrary complexity; and it is able to build on top of the experience of experts. The Fuzzy Logic inference system was built using the MATLAB toolbox; 3 fuzzy sets were described for each entry variable (Indicator) and 5 fuzzy sets for the output variable (Qualification). Both, the set of membership function and the set of rules were defined in combination. The methodology is simple but at the same time is powerful; it allows the use of all the indicators with their own magnitudes and units. Five reactors were compared: the Advanced Boiling Water Reactor (ABWR), the Economic Simplified Boiling Water Reactor (ESBWR), the Evolutionary Pressurized water Reactor (EPR), the Advanced Pressurized water reactor 1000 (AP1000) and the Pebbled Bed Modular Reactor (PBMR). Preliminary results were obtained using non official data obtained from public information. The qualifications of the reactors appear to be quite near. This work should be improved by taking into account which indicator is important and grading the indicators according to the situation in Mexico. (authors)

  8. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

    2005-01-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

  9. The influence of thorium on the temperature reactivity coefficient in a 400 MWth pebble bed high temperature plutonium incinerating reactor

    International Nuclear Information System (INIS)

    Richards, Guy A.; Serfontein, Dawid E.

    2014-01-01

    This article investigates advanced fuel cycles containing thorium and reactor grade plutonium (Pu(PWR)) in a 400 MW th Pebble Bed Modular Reactor (PBMR) Demonstration Power Plant. Results presented were determined from coupled neutronics and thermo-hydraulic simulations of the VSOP 99/05 diffusion codes. In a previous study impressive burn-ups (601 MWd/kg heavy metal (HM)) and thus plutonium destruction rates (69.2 %) were obtained with pure plutonium fuel with mass loadings of 3 g Pu(PWR)/fuel sphere or less. However the safety performance was poor in that the limit on the maximum fuel temperature during equilibrium operation was exceeded and positive Uniform Temperature Reactivity Coefficients (UTCs) were obtained. In the present study fuel cycles containing mixtures of thorium and plutonium achieved negative maximum UTCs. Plutonium only fuel cycles also achieved negative maximum UTCs, provided that much higher mass loadings are used. It is proposed that the lower thermal neutron flux was responsible for this effect. The plutonium only fuel cycle with 12 g Pu(PWR)/fuel sphere also achieved the adopted safety limits for the PBMR DPP-400 in that the maximum fuel temperature and the maximum power density did not exceed 1130°C or 4.5 kW/sphere respectively. This design would thus be licensable and could potentially be economically feasible. However the burn-up was much lower at 181 MWd/kgHM and thus the plutonium destruction fraction was also much lower at 24.5%, which may be sub-optimal with respect to proliferation and waste disposal objectives and therefore further optimisation studies are proposed. (author)

  10. CFD applications in the Pebble Bed Modular Reactor Project: A decade of progress

    International Nuclear Information System (INIS)

    Janse van Rensburg, J.J.; Kleingeld, M.

    2011-01-01

    Highlights: → This paper evaluates the evolution of Gas Cooled Reactor CFD analysis over the last decade. → It discusses the influence of advances in hardware and software on the evolution of capabilities. → The advances in mesh generation and the physics that can be included is also discussed. → The focus was on the capabilities rather than improving the assumptions and correlations. - Abstract: Of all the systems and components that have to be designed for a nuclear plant, the Reactor Unit is the most significant since it is at the very heart of the plant. At Pebble Bed Modular Reactor (Pty) Ltd. (PBMR), the design of the Reactor Unit is conducted with the aid of extensive analysis work. Due to the rapid computational improvements, the analysis capabilities have had to evolve rather significantly over the last decade. This paper evaluates the evolution of RU Computational Fluid Dynamics (CFD) analysis in particular and presents a historical timeline of the analyses conducted at PBMR. The influence of advances in the hardware and software applications on the evolution of the analysis capabilities is also discussed. When evaluating the evolution of analysis, it is important to look not only at the advances in mesh generation and the representation of the geometry, but also at the improvements regarding the physics that were included in the models. The discussion evaluates the improvements from the pre-conceptual analyses, the concept design, the basic design and finally, the detail design. It is however important to note that the focus of this research was on establishing a methodology for the integrated CFD analysis of High Temperature Reactors. It is recognized however that results from this research can currently only be used to investigate and understand trends and behaviors rather than absolute values. It was therefore required to also launch an extensive V and V program of which the focus was to verify the approach and validate the methodology that

  11. Modelo de Sistema Basado en Conocimiento en el Dominio de la Seguridad de Aplicaciones

    Directory of Open Access Journals (Sweden)

    María Victoria Bajarlía

    2013-12-01

    Full Text Available El objetivo es proponer un modelo de un sistema basado en conocimiento (SBC aplicado al análisis de seguridad de aplicaciones de gestión. El modelo se fundamenta en un sistema basado en conocimiento (SBC que cuenta con un componente cognitivo que le permite incorporar conocimiento. En virtud de que las amenazas y los ataques informáticos representan un problema constante y creciente se puede suponer que el SBC, a través del aprendizaje dinámico que lo mantendrá actualizado, podrá asistir a los especialistas en Seguridad de la Información, en el área de competencia, a la elaboración de Especificación de Requerimientos.

  12. Desarrollo de materiales compuestos avanzados basados en fibras de carbono para la industria aeroespacial

    OpenAIRE

    Rodriguez, Exequiel Santos

    2015-01-01

    En este trabajo se estudia el desarrollo de materiales compuestos de matriz polimérica reforzados con fibras de carbono para aplicaciones que presentan solicitaciones severas. Este es el caso de la industria aeroespacial, que utiliza componentes en cohetes y aeronaves que se ven sometidos a altas solicitaciones mecánicas y están expuestos a las altas temperaturas. Es por ello que se estudiaron materiales para diversas aplicaciones basados en fibras de carbono: preimpregnados de fibras y resin...

  13. Materiales nanocumpuestos basados en LDPE relleno con nanotubos de carbono con potenciales propiedades bactericidas

    OpenAIRE

    Benigno Escribano, Erika

    2015-01-01

    En este trabajo, se ha seleccionado como material de estudio polietileno de baja densidad, LDPE, ya que es un polímero con múltiples aplicaciones en diversos campos. Se busca como objetivo principal, preparar y caracterizar nuevos materiales basados en LDPE con potenciales propiedades antibacterianas. Para ello, se van a estudiar dos posibles maneras de conseguirlo: La primera de ellas, consiste en realizar un procesado mecánico sobre el polietileno, concretamente, una molienda del alta en...

  14. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  15. Use of virtual environments to reduce the construction costs of the next generation nuclear power reactors

    International Nuclear Information System (INIS)

    Whisker, V.E.; Baratta, A.J.

    2007-01-01

    The near term deployment of the next generation of reactors will only be successful if they are built on time and without the costly overruns experienced in the previous generation. One critical factor in achieving these goals is to ensure the design is optimized for constructability. In this work the authors explored the effectiveness of full-scale virtual reality simulation in the optimization of the design and construction of the next generation of nuclear reactors. The research tested the suitability of immersive virtual reality display technology in aiding engineers in evaluating potential cost reductions that can be realized by the optimization of design and installation and construction sequences. The intent of this research is to see if this type of technology can be used in capacities similar to those currently filled by full-scale physical mockups and desktop simulations. Using a fully-immersive five sided virtual reality system, known as a CAVE, the authors constructed a series of virtual mockups that represented two next generation nuclear power plants, the Westinghouse AP-1000 and the Pebble Bed Modular Reactor (PBMR). These virtual mockups were then tested as a design tool to help locate and correct problem areas, to optimize the construction sequence, and to assist with familiarizing trades people with the performance of maintenance activities. A series of experiments were performed to assess the usefulness of these virtual mockups in accomplishing these tasks. (authors)

  16. Three-Dimensional Analysis of the Hot-Spot Fuel Temperature in Pebble Bed and Prismatic Modular Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Lee, S. W.; Lim, H. S.; Lee, W. J.

    2006-01-01

    High temperature gas-cooled reactors(HTGR) have been reviewed as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor(PBR) and a prismatic modular reactor(PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

  17. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

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

  18. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  19. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  20. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  1. Control de inventarios aplicado a empresas comerciales basado en el Coso Erm

    OpenAIRE

    Alvear Zamora, Gladys Elizabeth

    2014-01-01

    Esta investigación tiene como objetivo realizar el control de inventarios de tres empresas comerciales de tipo familiar basado en el COSO ERM, esta es una herramienta que nos ayudará a detectar riesgos de fraude, robo y pedidos excesivos de productos. A su vez, se analizará la eficiencia del manejo del control de inventarios, por medio de los métodos o procedimientos para la toma de los mismos, sus saldos de inventarios (valores máximos y mínimos), porcentajes de ventas, ratios financieros qu...

  2. El etiquetado de la carne de vacuno basado en la trazabilidad del producto

    OpenAIRE

    Bravo, Ana

    2002-01-01

    A estas alturas decir, que debido a la crisis de las "vacas locas" se ha tomado conciencia de lo importante que es conocer la trazabilidad de los productos alimenticios parece obvio, pero no conviene olvidarnos de ello para no detener el avance en la identificación, registro y etiquetado de todos los productos agrícolas y alimenticios. Es muy importante el nuevo concepto, introducido por la normativa de etiquetado basado en la trazabilidad del producto, que es el de la “individualización de r...

  3. Aprendizaje basado en problemas, como potencializador del pensamiento matemático

    OpenAIRE

    López Ordoñez, Jairo; Hidalgo Paredes, Hernán Darío; Mera Gutiérrez, Eduardo Andrés; Patiño Giraldo, Luz Elena

    2015-01-01

    Artículo (Maestria en Educación desde la Diversidad). Universidad de Manizales. Facultad de Ciencias Sociales y Humanas, 2015 La investigación tuvo como propósito determinar la incidencia de la estrategia didáctica del Aprendizaje Basado en Problemas (ABP), en el mejoramiento de los resultados obtenidos de las pruebas SABER11 en el área de matemáticas, teniendo como sujetos a los educandos de la Institución Educativa “JORGE VILLAMIL CORDOVÉZ” de Pitalito (Huila). Esta estrategia didáctica...

  4. Vectores recombinantes basados en el virus Vaccinia modificado de Ankara (MVA) como vacunas contra la leishmaniasis

    OpenAIRE

    Pérez Jiménez, Eva; Larraga, Vicente; Esteban, Mariano

    2005-01-01

    Vectores recombinantes basados en el virus vaccinia modificado de Ankara (MVA) como vacunas contra la leishmaniasis. Los vectores de la invención contienen secuencias codificantes de la proteína LACK, preferentemente insertadas en el locus de hemaglutinina del virus y bajo el control de un promotor que permite su expresión a lo largo del ciclo de infección del virus. Son vectores seguros, estables, que dan lugar a una potente respuesta inmune que confiere protección frente a la leishmaniasis,...

  5. Implementación de un sistema de control de acceso basado en reconomiento facial

    OpenAIRE

    Poveda, Martín; Merchán, Fernando

    2016-01-01

    "En este artículo se presentan aspectos de la implementación de un sistema de control de acceso basado en reconocimiento facial. El mismo verifica en tiempo real si las personas que entran a las instalaciones forman parte de la base de datos del personal que labora en las mismas. Las condiciones de operación de este sistema son no colaborativas. Es decir, los usuarios no se ubicarán en posiciones específicas para la adquisición de las imágenes. Para solventar las dificultades que esto pued...

  6. Mantenimiento industrial basado en la gestión del conocimiento

    OpenAIRE

    Cárcel Carrasco, Francisco Javier

    2014-01-01

    El mantenimiento industrial requiere de conocimientos técnicos muy específicos, normalmente almacenados de manera tácita entre el personal que opera en estas áreas. Mediante la reseña sobre dos libros de investigación, en este artículo se muestra el nivel estratégico que para las empresas con activos físicos supone una adecuada gestión del conocimiento en la ingeniería del mantenimiento industrial. Cárcel Carrasco, FJ. (2014). Mantenimiento industrial basado en la gestión del conocimiento....

  7. Un modelo de recuperación de información basado en SVMs

    Directory of Open Access Journals (Sweden)

    M. Fernanda Maldonado

    2004-01-01

    Full Text Available Los clasificadores como los SVMs (Support Vector Machines se usaron para la clasificación de documentos de manera muy eficiente, pero su utilidad no ha sido comprobada para la recuperación de información (RI en el momento de jerarquerizar los documentos. En este artículo proponemos una transformación que asocia el proceso de la RI a un nuevo espacio vectorial en el que un clasificador basado en SVMs se entrena para aprender el concepto de similitud frente a los documentos.

  8. Coordinación de grupos en juegos móviles basados en posicionamiento

    OpenAIRE

    Inafuku, Fernando Gabriel; Galella, Pablo Fernando

    2017-01-01

    Coordinar tareas para ser llevadas a cabo por un grupo de personas permite obtener una acción unificada y uniforme donde la suma de los esfuerzos individuales se potencian en pos de resolver actividades. En el caso de los Juegos Móviles basados en Posicionamiento, requiere que los participantes se coordinen en posiciones (o áreas) determinadas del ambiente físico para poder avanzar en el juego. Se tomó como base la tesina de grado de Matías Apezteguía y Darío Rapetti denominada “Juego Educati...

  9. Sistema de riego autónomo basado en la Internet de las Cosas

    OpenAIRE

    Castro-Silva, Juan Antonio

    2016-01-01

    Las necesidades de agua para la producción de alimentos seguirán en aumento debido al crecimiento de la población mundial. En este trabajo de investigación se ha construido un sistema de riego autónomo basado en la Internet de las Cosas (IoT). Se emplean elementos de bajo costo y hardware - software libre (Raspberry Pi, Arduino, Linux, Java, Wildfly, Python, etc.) para implementar Redes de Sensores Inalámbricos (WSN) que permiten obtener la información de las variables agroclimáticas (Humedad...

  10. Sistema de costos basado en actividades en hoteles cuatro estrellas del estado Mérida, Venezuela

    OpenAIRE

    Marysela Coromoto Morillo Moreno; Cororina del Carmen Cardozo

    2017-01-01

    Las empresas actualmente demandan sistemas de costos que reporten un mayor detalle en la información generada, con el propósito de orientar la aplicación de estrategias que conduzcan a captar y apropiarse de mayores espacios de mercado, sobre todo cuando la competitividad es elevada; por ello, se formuló un sistema de costos basado en actividades, conocido por sus siglas en inglés como abc (Activity-Based Costing), aplicado a los hoteles de turismo de cuatro estrellas de Mérida, en Vene...

  11. Pebble bed modular reactor fuel enrichment discrimination using delayed neutrons - HTR2008-58133

    International Nuclear Information System (INIS)

    Skoda, R.; Rataj, J.; Uhera, J.

    2008-01-01

    The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it's on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method. (authors)

  12. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavron, Barak, E-mail: btavron@bgu.ac.il [Planning, Development and Technology Division, Israel Electric Corporation Ltd., P.O. Box 10, Haifa 31000 (Israel); Shwageraus, Eugene, E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

    2016-10-15

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  13. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    International Nuclear Information System (INIS)

    Tavron, Barak; Shwageraus, Eugene

    2016-01-01

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  14. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  15. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  16. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  17. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

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

  18. MODELO PARA EL PERFECCIONAMIENTO DE LAS COMPETENCIAS DEL INGENIERO INDUSTRIAL BASADO EN LABORATORIOS DE APRENDIZAJE

    Directory of Open Access Journals (Sweden)

    Eduyn Ramiro Lopez Santana

    2015-09-01

    Full Text Available Se propone un modelo basado en la metodología de dinámicas de sistemas que valida las bondades que trae al perfeccionamiento de las competencias del ingeniero industrial de la Universidad Distrital (Colombia basado en los principios de los laboratorios de aprendizaje. Su campo de aplicación permite re-significar el papel que tiene el estudiante y docente en el proceso de aprendizaje. El modelo parte del principio de causalidad que intrínsecamente representa el proceso de aprendizaje y como se puede reforzar el mismo a través de la práctica de la simulación como mecanismo de un laboratorio de aprendizaje. El resultado ha sido la elaboración de varios prototipos de modelo en donde el estudiante se enfrenta a casos de la vida real pero en un ambiente simulado. Los resultados arrojan una serie de propuestas que validan las premisas del modelo de perfeccionamiento, y por otro lado, permitirá en un futuro el desarrollo de modelos de simulación en el contexto de la concepción del laboratorio de aprendizaje.

  19. ¿Qué pueden ofrecer los modelos basados en agentes vivos en el contexto docente?

    Directory of Open Access Journals (Sweden)

    Marta Ginovart Gisbert

    2015-07-01

    Full Text Available Los sistemas biológicos o sistemas formados por entidades vivas (individuos son sistemas complejos, tanto por la “complejidad” que cada individuo o agente vivo tiene, como por las posibles relaciones que se pueden establecer entre ellos, así como por las posibles relaciones con el entorno o medioambiente en el que estos individuos se desarrollan, viven, compiten y mueren, y que por tanto, modifican como resultado de sus acciones. Este trabajo se basa en la experiencia acumulada en los últimos años en el uso de modelos basados en agentes en el ámbito de los biosistemas en la Universidad Politècnica de Catalunya. El objetivo es ofrecer elementos de estudio y discusión para poder responder a las siguientes preguntas: 1 ¿Qué son los modelos basados en agentes vivos?, 2 ¿Cómo se puede trabajar con estos modelos computacionales en el aula?, y 3 ¿Qué pueden ofrecer estos modelos en un entorno educativo? Asimismo, se proporciona información y referencias específicas para facilitar la incorporación de este tipo de modelo en planes de estudios con diferentes niveles de instrucción matemática y biológica, como complemento a otras metodologías de modelización.

  20. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  1. Multiregion reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

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

  3. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  4. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  5. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  6. Towards Compact Antineutrino Detectors for Safeguarding Nuclear Reactors

    International Nuclear Information System (INIS)

    Meijer, R.J. de; Smit, F.D.; Woertche, H.J.

    2010-01-01

    In 2008 the IAEA Division of Technical Support convened a Workshop on Antineutrino Detection for Safeguards Applications. Two of the recommendations expressed that IAEA should consider antineutrino detection and monitoring in its current R and D program for safeguarding bulk-process reactors, and consider antineutrino detection and monitoring in its Safeguards by Design approaches for power and fissile inventory monitoring of new and next generation reactors. The workshop came to these recommendations after having assessed the results obtained at the San Onofre Nuclear Generator Station (SONGS) in California. A 600 litre, 10% efficiency detector, placed at 25m from the core was shown to record 300 net antineutrino events per day. The 2*2.5*2.5 m 3 footprint of the detector and the required below background operation, prevents an easy deployment at reactors. Moreover it does not provide spatial information of the fissile inventory and, because of the shape of a PBMR reactor, would not be representative for such type of reactor. A solution to this drawback is to develop more efficient detectors that are less bulky and less sensitive to cosmic and natural radiation backgrounds. Antineutrino detection in the SONGS detector is based on the capture of antineutrinos by a proton resulting in a positron and neutron. In the SONGS detector the positron and neutron are detected by secondary gamma-rays. The efficiency of the SONGS detector is largely dominated by the low efficiency for gamma detection high background sensitivity We are investigating two methods to resolve this problem, both leading to more compact detectors, which in a modular set up also will provide spatial information. One is based on detecting the positrons on their slowdown signal and the neutrons by capturing in 10 B or 6 Li, resulting in alpha-emission. The drawback for standard liquid scintillators doped with e.g. B is the low flame point of the solvent and the strong quenching of the alpha signal. Our

  7. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  8. Solucionando dificultades en el aula: una estrategia usando el aprendizaje basado en problemas

    Directory of Open Access Journals (Sweden)

    Mery Luz Valderrama Sanabria

    2017-09-01

    Full Text Available Introducción: Se utiliza el proyecto de investigación en el aula con énfasis en el Aprendizaje Basado en Problemas, el cual promueve el aprendizaje activo y significativo, permitiendo solucionar situaciones reales de conocimiento en torno a una temática específica. Implementa los principios de la investigación formativa, como herramienta para generar nuevas alternativas en la apropiación del conocimiento. Este estudio tuvo como objetivo conocer la percepción de los estudiantes del programa Regencia de Farmacia frente a la utilización del aprendizaje basado en problemas con el fin de realizar aportes al currículo. Materiales y Métodos: Estudio descriptivo y transversal realizado con una muestra no probabilística, por conveniencia, conformada por 109 estudiantes de segundo a sexto semestre. Se elaboró un cuestionario con escala tipo Likert, sometido a valoración por expertos. Resultados: En general, los estudiantes están de acuerdo con la estrategia porque ha permitido acercarse a la investigación, fortaleciendo el pensamiento crítico; generando autonomía y responsabilidad frente al aprendizaje. A medida que avanzan los semestres, le ven mayor utilidad. Sin embargo, falta claridad en el uso de la metodología y capacitación por parte de algunos docentes para desarrollarla eficazmente. Discusión: La coordinación docente es fundamental, se debe fortalecer este aspecto para dar claridad al uso de la metodología. Implementar procesos de evaluación para determinar los avances desarrollados por los estudiantes y el impacto generado. Conclusiones: Los estudiantes consideran que adquieren conocimientos y competencias que les ayudarán en la práctica profesional.  Cómo citar este artículo: Valderrama ML, Castaño GA. Solucionando dificultades en el aula: una estrategia usando el aprendizaje basado en problemas. Rev Cuid. 2017; 8(3: 1907-18. http://dx.doi.org/10.15649/cuidarte.v8i3.456

  9. An endothermic chemical process facility coupled to a high temperature reactor. Part II: Transient simulation of accident scenarios within the chemical plant

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Revankar, Shripad T.

    2012-01-01

    Highlights: ► Seven quantitative transient case studies were analyzed in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Positive power excursion in the nuclear reactor were found for helium-inlet overcoolings. ► In all cases studied the maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Transient study of the operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven quantitative transient case studies are analyzed. The case studies consist of: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. Various parametric studies based on the magnitude of the events were also performed. The only chemical plant initiated events that caused a positive power excursion in the nuclear reactor were helium-inlet overcoolings due to process holding tank failures or reaction chamber ruptures. Even for a severe sustained overcooling, the calculated maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. The qualitative basis for the case studies and the analysis models are summarized in part I of this paper.

  10. Videojuego de rol táctico de estética retro basado en combates por turnos

    OpenAIRE

    García García, Juan de la Cruz

    2010-01-01

    Videojuego de rol basado en combates por turnos con estética retro, de nombre comercial TierraCuadrada, donde el jugador podrá llevar a cabo partidas en modo historia, involucrando combates con la Inteligencia Artificial, y partidas contra rivales humanos.

  11. Neutronic modeling of pebble bed reactors in APOLLO2

    International Nuclear Information System (INIS)

    Grimod, M.

    2010-01-01

    In this thesis we develop a new iterative homogenization technique for pebble bed reactors, based on a 'macro-stochastic' transport approximation in the collision probability method. A model has been developed to deal with the stochastic distribution of pebbles with different burnup in the core, considering spectral differences in homogenization and depletion calculations. This is generally not done in the codes presently used for pebble bed analyses, where a pebble with average isotopic composition is considered to perform the cell calculation. Also an iterative core calculation scheme has been set up, where the low-order RZ S N full-core calculation computes the entering currents in the spectrum zones subdividing the core. These currents, together with the core k eff , are then used as surface source in the fine-group heterogeneous calculation of the multi-pebble geometries. The developed method has been verified using reference Monte Carlo simulations of a simplified PBMR- 400 model. The pebbles in this model are individually positioned and have different randomly assigned burnup values. The APOLLO2 developed method matches the reference core k eff within ± 100 pcm, with relative differences on the production shape factors within ± 4%, and maximum discrepancy of 3% at the hotspot. Moreover, the first criticality experiment of the HTR-10 reactor was used to perform a first validation of the developed model. The computed critical number of pebbles to be loaded in the core is very close to the experimental value of 16890, only 77 pebbles less. A method to calculate the equilibrium reactor state was also developed and applied to analyze the simplified PBMR-400 model loaded with different fuel types (UO 2 , Pu, Pu + MA). The potential of the APOLLO2 method to compute different fluxes for the different pebble types of a multi-pebble geometry was used to evaluate the bias committed by the average composition pebble approximation. Thanks to a 'compensation of error

  12. Oscilador para biosensores basado en microbalanza de cristal de cuarzo (QCM

    Directory of Open Access Journals (Sweden)

    Yeison Javier Montagut Ferizzola

    2011-01-01

    Full Text Available El cristal de cuarzo generalmente es usado en aplicaciones como microbalanza, aprovechando la capacidad que presenta éste para variar su frecuencia de resonancia de acuerdo a los cambios de la densidad superficial de masa depositada en la superficie del resonador. De esta manera, un cristal de cuarzo puede ser utilizado como transductor en un sistema de inmunosensor piezoeléctrico, para detectar uniones antígeno - anticuerpo. En este artículo se presenta una interfaz para microbalanzasde cristal de cuarzo, QCM (del inglés Quartz Crystal Microbalance basado en una versión mejorada de oscilador en configuración diferencial equilibrado y su validación como sistema de caracterización para biosensores. El sistema fue probado con éxito en un inmunosensor piezoeléctrico para la detección del plaguicida Carbaryl.

  13. Ambientes y diseño de escenarios en el aprendizaje basados en simulació

    Directory of Open Access Journals (Sweden)

    MSc. Betty Bravo Zúñiga

    2018-01-01

    Full Text Available Se aborda nuevas metodologías educativas que guíen al docente y estudiantes a la reflexión de su praxis mediante el debriefing y el feedback, mediante la observación de grabaciones o registros realizados durante la práctica, las cuales generan discusiones y permiten evaluar el desempeño de los casos clínicos simulados. Se concluye que el aprendizaje basado en simulación puede ser optimizado mediante el diseño de guías, planificación y evaluación de los escenarios y la implementación de los affordance en el ambiente

  14. Modelos de Requisitos Basados en I* para Detectar Proactividad en Dashboards

    Directory of Open Access Journals (Sweden)

    Alain Pérez-Acosta

    2014-06-01

    Full Text Available El objetivo del trabajo es presentar modelos para la captura de los requisitos de un dashboard para detectar un comportamiento proactivo. Estos modelos siguen un enfoque orientado hacia metas y fueron creados con el marco de trabajo i*, que toma como base las premisas del modelado social. Para detectar el comportamiento proactivo se usaron patrones basados en modelos de i* para detectar proactividad en la etapa de requisitos de un sistema de software. Los modelos que se obtienen como resultado del trabajo tienen representados los actores, metas, intenciones, tareas y recursos que se necesitan para modelar los requisitos de un dashboard con un comportamiento proactivo y pueden, además, ser utilizados en distintos contextos de negocio.

  15. Satisfacción Laboral: Un Modelo Explicativo Basado en Variables Disposicionales

    Directory of Open Access Journals (Sweden)

    Solana Magali Salessi

    2017-07-01

    Full Text Available Se verificó un modelo explicativo de la satisfacción laboral basado en la relación de algunas variables disposicionales. Se plantea un modelo de mediación múltiple moderada, que fue analizado en una muestra multiocupacional de 575 trabajadores argentinos. La verificación empírica indicó que el efecto indirecto positivo del capital psicológico y de la inteligencia emocional sobre la satisfacción laboral es amplificado por la extraversión y atenuado por el neuroticismo; en contraste, el efecto indirecto negativo del cinismo organizacional se encuentra fortalecido por el descontrol emocional y amortiguado por la extraversión. Se discuten los resultados y se señalan las fortalezas y limitaciones del estudio.

  16. Control Servo-Visual de un Robot Manipulador Planar Basado en Pasividad

    Directory of Open Access Journals (Sweden)

    Carlos Soria

    2008-10-01

    Full Text Available Resumen: En este trabajo se diseña un controlador servo visual basado en la propiedad de pasividad del sistema visual. Se propone un regulador con ganancias de control variables, de tal manera que se evita la saturación de los actuadores y al mismo tiempo presenta la capacidad de corregir errores de pequeña magnitud. Asimismo el diseno se hace tenieñdo en cuenta el desempeño L2, a fin de darle capacidad de seguimiento de objetos en movimiento, con un error de control pequeño. Se muestran resultados experimentales realizados en un robot manipulador industrial tipo planar para verificar el cumplimiento de los objetivos del controlador propuesto. Palabras Clave: robot manipulador industrial, control servo visual, control no lineal, pasividad

  17. Mantenimiento Correctivo Aplicado a un Sitio Basado en Joomla. Una Propuesta Centrada en la Accesibilidad

    Directory of Open Access Journals (Sweden)

    Daiana E. Casaro Pedro L. Alfonzo

    2015-05-01

    Full Text Available Se presenta una experiencia de personalización de un sitio basado en un sistema gestor de contenidos de libre distribución o CMS, atendiendo su amplia aplicabilidad en la Industria del Software y el Sector de Servicios Informáticos. Dado que un CMS puede ser mejorado por la comunidad de desarrolladores, se considera al mantenimiento correctivo como parte integral del ciclo de vida de desarrollo del software. En este trabajo, se abordan los estándares propuestos por el Consorcio W3C que permiten el acceso a los contenidos web y la verificación del cumplimiento de las pautas WCAG 2.0, durante el proceso de desarrollo del sitio web.

  18. Perfiles de personas con diferencia mental basados en las funciones motrices gruesas

    Directory of Open Access Journals (Sweden)

    Isabel Ma. Ferrándiz Vindel

    2002-01-01

    Full Text Available El propósito de este estudio ha sido identificar posibles subtipos de alumnos con Deficiencia Mental basados en funciones motrices gruesas. Se seleccionó un grupo de alumnos de un centro específico de Deficiencia Mental, que no presentaron deficiencias motóricas asociadas, y se les administraron los subtests de motricidad gruesa del «Test Bruininks-Ozeretsky para medir la eficacia motriz»- Los cuatro subtipos resultantes mostraron diferentes perfiles de desarrollo motor A partir de estos resultados, se recomienda diseñar programas específicos que intenten paliar los trastornos motrices gruesos, analizados para cada uno de los subtipos hallados, y realizar adaptaciones curriculares pertinentes, teniendo en cuenta los perfiles motrices estudiados

  19. Valoración del profesorado de magisterio sobre el aprendizaje basado en competencias implantado

    Directory of Open Access Journals (Sweden)

    Aurelio Villa

    2013-01-01

    Full Text Available A partir del curso académico 2010 - 11 se ha iniciado en Europa de modo oficial el Espacio Europeo de Educación Superior (EEES, y con él la puesta en marcha de las denominadas titulaciones Bolonia bajo la etiqueta del crédito europeo (European Credit Transf er System, ECTS. Este artículo trata de presentar el nivel de desarrollo del aprendizaje basado en competencias de las titulaciones de magisterio en España. A través de una metodología denominada Análisis Importancia - Realización (AIR (Martilla, J.A. y J ames, J.L., 1977 con la que se pretende dibujar la valoración que el profesorado y responsables académicos de magisterio realizan sobre el aprendizaje basado en competencias (ABC, y extraer las principales conclusiones del estudio empírico llevado a cabo en una muestra de 145 personas (profesores y gestores. Se presentan los resultados de un análisis factorial con cinco factores que describen el proceso de enseñanza - aprendizaje: planificación de competencias, gestión pedagógica, coordinación docente, tut oría y evaluación, y finalmente, revisión y mejora. Estos factores se correlacionaron con la formación recibida para iniciar el EEES y con el uso de la lección magistral. Obteniendo ambas variables correlaciones significativas. Este artículo se enriquece con otro artículo complementario que recoge la perspectiva cualitativa del profesorado sobre su visión de la implementación del proceso europeo y sus dificultades.

  20. El currículo basado en competencias profesionales integradas en la universidad ecuatoriana

    Directory of Open Access Journals (Sweden)

    Floralba del Rocío Aguilar Gordón

    2017-01-01

    Full Text Available El artículo reflexiona acerca del currículo basado en competencias profesionales integradas en las carreras de educación de la universidad ecuatoriana, analiza los principios y fundamentos que orientan al quehacer educativo representado por una diversidad de concepciones ideológicas, antropológicas, sociológicas, epistemológicas, pedagógicas y psicológicas en las que se concretizan. Consta de cuatro partes: En la primera parte, presenta los antecedentes del tema desarrollado; aborda los principales problemas, necesidades de los contextos y objetivos; determina las tendencias de desarrollo local y regional incluidas en los campos de estudio; plantea los principales requerimientos de la sociedad ecuatoriana. En la segunda parte, analiza los principios generales que rigen el currículo en base a competencias profesionales, se refiere a los campos de formación del currículo; a los niveles de organización curricular; a los núcleos básicos de las disciplinas que sustentan la profesión. En la tercera parte, reflexiona acerca de la planificación curricular por competencias; revisa las orientaciones del conocimiento y los saberes que tiene en cuenta la construcción del objeto de estudio de la profesión. En la cuarta parte, presenta algunas contribuciones del currículo basado en competencias profesionales integradas para la formación profesional y para la formación del talento humano.

  1. Control en red basado en eventos: de lo centralizado a lo distribuido

    Directory of Open Access Journals (Sweden)

    María Guinaldo

    2017-01-01

    Full Text Available Resumen: Los sistemas de control en red (SCR son aquellos en los que los diferentes elementos de un lazo de control (sensores, actuadores y controladores se encuentran espacialmente distribuidos y la transmisión entre ellos tiene lugar a travón de informaciés de un canal de comunicación o red. La reducción de la cantidad de información transmitida juega un papel importante en el desempeño de estos sitemas, y reglas de comunicación no convencionales como el control basado en eventos, se han demostrado efectivas. En este artículo se revisan algunas de estas estrategias, centrándose en primer lugar en los SCR centralizados, para posteriormente estudiar esquemas de control distribuido, aplicados a sistemas de gran escala. Finalmente, algunos de los resultados teóricos se aplican al control de formaciones en un sistema de experimentación real. Abstract: Networked control systems (NCSs are spatially distributed systems in which sensors, actuators and controllers exchange information through a communication channel or network. The reduction in the amount of transmitted information has a relevant impact over the system's performance. In this regard, non-conventional communication rules, such as the event-based control, have been demonstrated to be effective. In this paper, some of these strategies are reviewed. We first focus on centralized NCSs, and then the distributed control for large-scale systems is studied. Finally, some of the results are applied to the formation control problem and implemented over an experimental setup. Palabras clave: Control en red, control basado en eventos, control distribuido, sistemas de gran escala, sistemas multi-agente, control de formaciones, robots móviles., Keywords: Networked control, event-based control, distributed control, large-scale system, multi-agent system, formation control, mobile robot

  2. Aprendizaje basado en la resolución de problemas: una experiencia práctica

    Directory of Open Access Journals (Sweden)

    E. González-López

    Full Text Available El aprendizaje basado en la resolución de problemas incorpora herramientas metodológicas capaces de facilitar la consecución de los objetivos propuestos para la formación de los futuros médicos dentro del marco de la docencia universitaria en el Espacio Europeo de Educación Superior. Promueve una formación más activa, flexible y práctica, que concede mayor protagonismo al trabajo personal tutorizado (aprendizaje autodirigido, en detrimento de las clásicas clases teóricas, eminentemente expositivas, en las que el papel del estudiante es, en general, más pasivo. La Unidad de Medicina de Familia de la Universidad Autónoma de Madrid incorporó el aprendizaje basado en la resolución de problemas en el desarrollo de la asignatura optativa 'Atención Primaria y Medicina de Familia', ofertada como optativa a los alumnos de segundo ciclo de licenciatura (cursos 4.º a 6.º desde el curso 2005-2006. Intentamos con ella promover la formación de médicos capaces de aprender y mantener su competencia durante toda su vida profesional, no sólo en lo referido a la adquisición/integración de conocimientos científicos suficientes, sino también en cuanto al desarrollo de las habilidades necesarias para su adecuada aplicación práctica considerando a cada paciente de modo integral como realidad biopsicosocial, en un contexto sanitario definido, sin olvidar los aspectos bioéticos implícitos al quehacer del médico (respeto hacia el paciente y compromiso social. Revisamos en este artículo el diseño práctico de la asignatura.

  3. SELECCIÓN DE NUEVOS GENOTIPOS DE ARROZ BASADOS EN LA PROBABILIDAD DE SUPERAR AL TESTIGO

    Directory of Open Access Journals (Sweden)

    Ismael Camargo-Buitrago

    2014-01-01

    Full Text Available El objetivo de este trabajo fue validar una metodología estadística para estimar la confiabilidad o respuesta normalizada (RN i y la esta - bilidad de cuatro genotipos elite de arroz, en comparación con el testigo IDIAP 145-05. Se utilizó la base de datos del proyecto de mejoramiento genético de arroz del IDIAP, proveniente de los experimentos realizados entre el 2009 y 2011, en 31 ambientes bajo condiciones de secano. Los resultados del estudio permitieron verificar que los nuevos genotipos superaron significativamente (P<0,05 en rendi - miento al testigo. Los cuatro genotipos IDIAP FL 106-11, IDIAP FL 137-11, IDIAP FL 155, e IDIAP FL 156 presen - taron una confiablidad promedio de 0,79; 0,75; 0,75 y 0,74, respectivamente. La probabilidad normalizada del IDIAP FL 106-11 representó una respuesta diferencial en rendimiento mayor que cero con respecto al IDIAP 145-05, en ocho de cada diez casos. La confiabilidad estuvo relacionada con los parámetros de estabilidad basados en modelos de regresión (b i y S 2 di . El modelo multivariado AMMI, considerando el PCA1, identificó el genotipo IDIAP FL 156, como el más estable. El modelo Biplot GGE, basado en el PCA2, indicó que el genotipo IDIAP FL 155, tuvo mayor estabilidad. La confiabilidad o respuesta normalizada, puede ser útil para hacer recomendaciones más precisas para la utilización de los nuevos genotipos a nivel comercial

  4. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  5. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  7. Current status and future development of modular high temperature gas cooled reactor technology

    International Nuclear Information System (INIS)

    2001-02-01

    This report includes an examination of the international activities with regard to the development of the modular HTGR coupled to a gas turbine. The most significant of these gas turbine programmes include the pebble bed modular reactor (PBMR) being designed by ESKOM of South Africa and British Nuclear Fuels plc. (BNFL) of the United Kingdom, and the gas turbine-modular helium reactor (GT-MHR) by a consortium of General Atomics of the United States of America, MINATOM of the Russian Federation, Framatome of France and Fuji Electric of Japan. Details of the design, economics and plans for these plants are provided in Chapters 3 and 4, respectively. Test reactors to evaluate the safety and general performance of the HTGR and to support research and development activities including electricity generation via the gas turbine and validation of high temperature process heat applications are being commissioned in Japan and China. Construction of the high temperature engineering test reactor (HTTR) by the Japan Atomic Energy Research Institute (JAERI) at its Oarai Research Establishment has been completed with the plant currently in the low power physics testing phase of commissioning. Construction of the high temperature reactor (HTR-10) by the Institute of Nuclear Energy Technology (INET) in Beijing, China, is nearly complete with initial criticality expected in 2000. Chapter 5 provides a discussion of purpose, status and testing programmes for these two plants. In addition to the activities related to the above mentioned plants, Member States of the IWGGCR continue to support research associated with HTGR safety and performance as well as development of alternative designs for commercial applications. These activities are being addressed by national energy institutes and, in some projects, private industry, within China, France, Germany, Indonesia, Japan, the Netherlands, the Russian Federation, South Africa, United Kingdom and the USA. Chapter 6 includes details

  8. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  10. Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

    International Nuclear Information System (INIS)

    Shi, Dunfu

    2015-01-01

    The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

  11. An endothermic chemical process facility coupled to a high temperature reactor. Part I: Proposed accident scenarios within the chemical plant

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

    2012-01-01

    Highlights: ► The paper identifies possible transient and accident scenarios in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Key accidents scenarios were investigated through qualitative reasoning. ► The accidents were found to constitute loss of heat sink event for the nuclear reactor. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Quantitative study of the possible operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven unique case studies are proposed based on a thorough review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. The qualitative parameters of each case study are outlined as well as the basis in literature. A previously published modeling scheme is described and adapted for application as a simulation platform for these transient events. The results of the quantitative case studies are described within part II of this paper.

  12. Una adaptación del Proceso Unificado de Desarrollo para la creación de portales basados en Joomla

    Directory of Open Access Journals (Sweden)

    Enrique José Leyva Miranda

    2006-01-01

    Full Text Available Presenta una adaptación del Proceso Unificado (UP para la creación de portales basados en Joomla, un sistema de gestión de contenidos (CMS de código libre con características de plataforma para desarrollo de portales basado en componentes. En la modelación del proceso se ha empleado SPEM, un estándar basado en UML propuesto por el "Object Management Group" (OMG. Se proponen y describen cinco disciplinas e igual número de roles que se consideran adecuados a las características de este tipo de proyecto.

  13. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  14. Reactor safety

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  15. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  16. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  17. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

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

  18. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

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

  19. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

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

  20. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

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

  1. Reactor container

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

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

  3. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  4. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

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

  5. Sistema de Alerta al Conductor Basado en Realimentación Vibro-Táctil

    Directory of Open Access Journals (Sweden)

    Emanuel Slawiñski

    2015-01-01

    Full Text Available Resumen: Este trabajo propone el diseño y desarrollo de un sistema de alerta al conductor basado en la realimentación de estimulos vibro- táctiles de fuerza con el objetivo de prevenir accidentes de tránsito. El sistema posee dos agarres vibro-tactiles, los cuales se pueden montar facilmente sobre cualquier tipo de vehículo, y un sistema electrónico basado en un sistema de localización y comunicación inalámbrica entre vehículos, que permite calcular en línea una señal de alerta vibro-táctil para avisar al conductor de una posible situación de peligro en los proximos segundos. Un modelo focalizado en factores humanos es propuesto y utilizado para justificar el uso adecuado de estímulos artificiales. Además se describen, el hardware, la comunicación entre vehículos y software embebido. Finalmente, el sistema es probado en un simulador 3D de carrera de código abierto y también utilizando dos vehículos comunes. Abstract: This paper proposes the design and build of a driver warning system, based on vibro-tactile feedback for preventing accidents through the generation of tactile stimuli. The system has two vibro-tactile grips devices which are easily mounted on the steering wheel of any vehicle and an electronic system based on location sensing as well as inter-vehicles communication, from which a risk level is computed on line in order to warn the driver about dangerous situations and risk zones. A model focalized on human factors is proposed and it is employed to justify the advantages of using artificial stimuli. Besides, the hardware, communication between vehicles and embedded software, are described too. Finally, experiences using the device in a racing car simulator and tests using two ordinary cars are shown. Palabras clave: seguridad, sensores e instrumentos virtuales, automoción, estímulos táctiles de fuerza, prevención de accidentes, Keywords: Safety, sensors and virtual instrument, automotive, vibro

  6. Aprendizaje basado en programación. Alimento transdisciplinar para el pensamiento creativo

    Directory of Open Access Journals (Sweden)

    Alberto Domingo Galán

    2013-11-01

    Full Text Available Este trabajo describe una actividad de aprendizaje transdisciplinar destinado a impulsar la creatividad y fomentar actitudes innovadoras. Combina biología molecular con programación de computadoras, pero más que el caso concreto, es un ejemplo de abordaje dirigido a impulsar la capacidad de innovación en el proceso educativo. Se basa en la convicción de que un conocimiento amplio y multidisciplinar es el sustrato más eficiente para transformar la creatividad individual en innovación práctica. Esta actividad, ya aplicada en más de seis cursos, utiliza una metodología de aprendizaje activo inscrita en el marco del Aprendizaje Basado en Proyectos y se diseñó inicialmente para las clases prácticas de laboratorio de una asignatura de Biología Molecular. Los participantes tienen que desarrollar una herramienta de software real, capaz de realizar un análisis de secuencias de ácidos nucleicos, incluyendo una interfaz de usuario interactiva. Comienzan a escribir el código real y a aprender los rudimentos de un la programación directamente aplicándolos a resolver un problema. Se trata de un “aprendizaje basado en programación” ya que la escritura del programa no es la finalidad, sino el medio para crear un entorno en el que los alumnos tienen que repensar y entender realmente la lógica detrás de muchos conceptos biológicos fundamentales ya aprendidos, pero que en esta actividad tienen que enfrentar en un contexto diferente y con un nuevo lenguaje y objetivos prácticos. En lugar de descripciones teóricas, el proceso de aprendizaje involucra a los participantes en un reto divertido y motivador, una meta exigente pero estimulante de crear su propia solución para un problema auténtico, que se reconoce como difícil, pero también como comprensible y realizable. Es una actividad realmente transdisciplinar que también trata de dar a los estudiantes una visión de un nuevo campo, ampliando su mente y motivándoles a explorar y

  7. Profesionalización de promotores de lectura con el aprendizaje basado en proyectos mediado por TIC

    Directory of Open Access Journals (Sweden)

    Mario Miguel Ojeda-Ramírez

    2017-01-01

    Full Text Available Se revisa brevemente e l aprendizaje basado en proyectos (ABP mediado por TIC con el fin de motivar el contexto adecuado y los fundamentos para presentar el plan de estudios y las estrategias de aprendizaje utilizadas en la Especialización en Promoción de la Lectura (EPL, un programa de posgrado de la Universidad Veracruzana. Se describe e l uso de EMINUS, un entorno virtual de aprendizaje institucional , bajo un enfoque bimodal o blended learning; destacando el uso de la bitácora y los resultados asociados con el diseño, la ejecución y el reporte de los proyectos de promoción de la lectura. Se discuten los resultados de las dos primeras generaciones; p or último, se identifican las áreas de oportunidad de mejora de este sistema de entrenamiento basado en un esquema de aprendizaje personalizado med iado por TIC.

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

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

  9. Neutronic characterization of the SAFARI-1 material testing reactor - HTR2008-58155

    International Nuclear Information System (INIS)

    Makgopa, B. M.; Belal, M.; Strydom, W. J.

    2008-01-01

    This work presents a neutronic analysis of the core in the South African Fundamental Atomic Research Installation (SAFARI-1) for future Pebble Bed Modular Reactor (PBMR) fuel irradiation experiments. Monte Carlo simulation of the core with and without the rig has been performed. The results show a negligibly small reactivity worth of the rig, which is expected, due to the small amount of heavy metal loading in the pebble and the low fuel enrichment. This effect will be further investigated when the rig is extended to include more than one fuel pebble. Results further show perturbations in the neutron and photon flux as well as the power distribution in core position B6. A 50% thermal neutron flux depression is observed in position B6 due to the insertion of the rig. A 60% increase in axial photon heating values is also observed in position B6. The neutron and photon flux and power distributions in the other in-core irradiation positions (D6 and F6) are slightly affected by the insertion of this rig. Fluxes and power distributions in positions D6 and F6 will be studied in detail when they are loaded with isotope production rigs. (authors)

  10. Contact detection acceleration in pebble flow simulation for pebble bed reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y.; Ji, W. [Department of Mechanical, Aerospace, and Nuclear Engineering Rensselaer, Polytechnic Institute, 110 8th street, Troy, NY 12180 (United States)

    2013-07-01

    Pebble flow simulation plays an important role in the steady state and transient analysis of thermal-hydraulics and neutronics for Pebble Bed Reactors (PBR). The Discrete Element Method (DEM) and the modified Molecular Dynamics (MD) method are widely used to simulate the pebble motion to obtain the distribution of pebble concentration, velocity, and maximum contact stress. Although DEM and MD present high accuracy in the pebble flow simulation, they are quite computationally expensive due to the large quantity of pebbles to be simulated in a typical PBR and the ubiquitous contacts and collisions between neighboring pebbles that need to be detected frequently in the simulation, which greatly restricted their applicability for large scale PBR designs such as PBMR400. Since the contact detection accounts for more than 60% of the overall CPU time in the pebble flow simulation, the acceleration of the contact detection can greatly enhance the overall efficiency. In the present work, based on the design features of PBRs, two contact detection algorithms, the basic cell search algorithm and the bounding box search algorithm are investigated and applied to pebble contact detection. The influence from the PBR system size, core geometry and the searching cell size on the contact detection efficiency is presented. Our results suggest that for present PBR applications, the bounding box algorithm is less sensitive to the aforementioned effects and has superior performance in pebble contact detection compared with basic cell search algorithm. (authors)

  11. Contact detection acceleration in pebble flow simulation for pebble bed reactor systems

    International Nuclear Information System (INIS)

    Li, Y.; Ji, W.

    2013-01-01

    Pebble flow simulation plays an important role in the steady state and transient analysis of thermal-hydraulics and neutronics for Pebble Bed Reactors (PBR). The Discrete Element Method (DEM) and the modified Molecular Dynamics (MD) method are widely used to simulate the pebble motion to obtain the distribution of pebble concentration, velocity, and maximum contact stress. Although DEM and MD present high accuracy in the pebble flow simulation, they are quite computationally expensive due to the large quantity of pebbles to be simulated in a typical PBR and the ubiquitous contacts and collisions between neighboring pebbles that need to be detected frequently in the simulation, which greatly restricted their applicability for large scale PBR designs such as PBMR400. Since the contact detection accounts for more than 60% of the overall CPU time in the pebble flow simulation, the acceleration of the contact detection can greatly enhance the overall efficiency. In the present work, based on the design features of PBRs, two contact detection algorithms, the basic cell search algorithm and the bounding box search algorithm are investigated and applied to pebble contact detection. The influence from the PBR system size, core geometry and the searching cell size on the contact detection efficiency is presented. Our results suggest that for present PBR applications, the bounding box algorithm is less sensitive to the aforementioned effects and has superior performance in pebble contact detection compared with basic cell search algorithm. (authors)

  12. Modelos organizativos basados en el conocimiento. Desde la gerencia de la información a la gestión del conocimiento

    OpenAIRE

    Olmedo Narbona, Antonio

    2011-01-01

    Esta tesis doctoral estudia la evolución de los modelos de gestión del conocimiento, analizando la relación entre el modelo EFQM y los modelos organizativos basados en el conocimiento. Como aplicación práctica se propone el desarrollo de un modelo organizativo basado en el conocimiento para los Centros del Profesorado, unidades administrativas de la Junta de Andalucía.

  13. Algoritmo de reconocimiento de patrones basado en codificación fisiológica en cerebro de primates.

    OpenAIRE

    CASTEL BAIXAULI, ALEJANDRO

    2017-01-01

    El cerebro humano y animal es capaz de reconocer una gran cantidad de patrones, como por ejemplo caras, utilizando un limitado número de neuronas y algoritmos de procesamiento [Chang et al 2017]. Sin embargo, los algoritmos de reconocimiento de patrones utilizados en la actualidad, incluso aquellos basados en redes neuronales, requieren un alto número de operaciones y capacidad de computo. El objetivo del presente trabajo final de grado es adaptar, desarrollar y validar un algoritmo de re...

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

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

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

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

  16. Reactor container

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  17. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

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

  18. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  19. Modelo de Verificación y Validación Basado en CMMI

    Directory of Open Access Journals (Sweden)

    Osvaldo Puello

    2013-01-01

    Full Text Available La creciente preocupación por la calidad en la industria del software tiene como objetivo principal el desarrollo sistemático de productos y servicios de mejor calidad y el cumplimiento de las necesidades y expectativas de los clientes. En este artículo se presenta una introducción al modelo CMMI en sus niveles dos y tres. Seguidamente y basándose en la investigación “Metodología al proceso de verificación y validación de software basado en el estándar CMMI”, se plantea un modelo para verificar y validar software describiendo la planificación, actividades y estrategias para aplicar estos procesos durante el ciclo de vida del software.   Abstract The growing concern for quality in the software industry have as its main objective the systematic development of products and services of better quality and fulfilling the needs and expectations from customers. This article presents an overview to the CMMI model at levels 2 and 3. Then and on the basis of the "Methods A the process of verifying and validation of SOFTWARE based on the standard CMMI" research presents a model to verify and validate software describing planning, activities and strategies to apply these processes during the software life cycle.

  20. Los museos: un instrumento para el Aprendizaje Basado en Problemas (ABP

    Directory of Open Access Journals (Sweden)

    Yosajandi Pérez Campillo

    2011-01-01

    Full Text Available Diversos estudios han demostrado que el Aprendizaje Basado en Problemas (ABP es una propuesta educativa innovadora, que se caracteriza porque promueve que el aprendizaje sea significativo y contribuye a desarrollar una serie de habilidades y competencias indispensables para el crecimiento intelectual de cualquier persona. Sin embargo, esta estretegia implica repensar los problemas como problemas para aprender partiendo de preguntas que sean relevantes para los alumnos en el contexto del aprendizaje de ciencias. Y es justamente, el planteamiento de estas preguntas lo que hace complejo al ABP, pues platear "buenas preguntas" no es fácil, se requiere no sólo de habilidad y práctica sino también un conocimiento mínimo del tema y motivación para iniciar una investigación. Por lo anterior, una de las preocupaciones es encontrar los mecanismos que permitan introducir y motivar al estudiante para que sea capaz de plantear preguntas (problemas. En el presente trabajo se describen la propuesta de una serie de actividades que tienen como propósito preparar a un grupo de estudiantes de bachillerato para que planteen "buenas preguntas" de investigación sobre el tema de minerales. Para lograrlo, se considera el uso de varios recursos como: la visita a los museos, la lectura y análisis de textos y la actividad experimental.

  1. Tratamiento binocular de la ambliopía basado en la realidad virtual

    Directory of Open Access Journals (Sweden)

    Yanet Cristina Díaz Núñez

    Full Text Available Aunque los tratamientos predominantes de la ambliopía son monoculares, estos tienen poca aceptación y baja efectividad en el restablecimiento de la combinación binocular. Numerosas evidencias apoyan la idea de que la ambliopía es en esencia un problema binocular y que la supresión juega un papel clave. En esta revisión se exponen dos estrategias para el tratamiento binocular de la ambliopía basado en la realidad virtual; la primera con el objetivo primario de mejorar la agudeza visual y la segunda con el propósito de mejorar las funciones binoculares a través de la reducción de la supresión. Este enfoque binocular expone al paciente a condiciones artificiales de visión con estímulos dicópticos en imágenes relacionadas. Los estudios clínicos realizados, tanto en niños como adultos, reportan mejorías de la agudeza visual y la estereopsia en un tiempo muy inferior al requerido por la oclusión. Los resultados clínicos sugieren que un enfoque binocular que combine ambas estrategias puede utilizarse como complemento de los tratamientos clásicos y como alternativa en adultos y niños con historial de tratamientos fracasados o rechazados.

  2. Cuidados basados en narrativas: redefiniendo la jerarquía de la evidencia

    Directory of Open Access Journals (Sweden)

    Lorenzo Mariano Juárez

    2013-06-01

    Full Text Available Desde finales de los noventa proliferan trabajos que, aglutinados como Narrative Based Medicine (NBM, persiguen redefinir la práctica clínica orientándola hacia el paciente y su narrativa. Emergidos como un discurso eminentemente crítico y contestatario sobre la ortodoxia de la Evidence Based Medicine (EBM, actualmente persiguen perspectivas integracionistas que pretenden incluir sendos movimientos en un modelo de atención que reconstruya las nociones y jerarquía de la "evidencia". Ideológicamente, la NBM propone una redefinición de la ethos de la práctica médica resituando la experiencia del paciente como evidencia de primer orden. Sin embargo, el impacto de este movimiento en los cuidados enfermeros ha sido prácticamente nulo. Este artículo, primero de una serie, explora las bases teóricas y principios de la NBM sugiriendo un marco de trabajo que cimiente unos cuidados de Enfermería Basados en Narrativas, incluyendo una revisión de los estudios enmarcados en esta orientación y una reflexión sobre su carácter "vanguardista".

  3. Emulación en hardware de circuitos cuánticos basados en compuertas Toffoli

    Directory of Open Access Journals (Sweden)

    Jorge E. Duarte-Sánchez

    2014-01-01

    Full Text Available Este trabajo presenta el diseño de una arquitectura hardware para emular circuitos cuánticos basados en compuertas tipo Toffoli con la cual se pueden emular circuitos de más de 50 qubits. El estado del sistema se obtiene procesando independientemente cada estado base mediante las funciones determinadas por las compuertas cuánticas para cada qubit; el tiempo requerido para la emulación crece exponencialmente en función del número de qubits que se usan solamente para generar una superposición de estados, y no en función de la cantidad total de qubits del sistema como ocurre cuando se usa la representación matricial convencional. Adicionalmente, se diseñó un arreglo de unidades de procesamiento para disminuir el tiempo de ejecución. Los resultados de síntesis permiten concluir que se requieren 9,35 segundos para emular la exponenciación modular de 8 bits, la cual utiliza 48 qubits, 155.312 compuertas cuánticas y requiere procesar 131.072 estados base. Estos resultados también permiten estimar que se pueden implementar 256 unidades de procesamiento de 52 qubits en el FPGA EP3C120F780I7.

  4. USO DEL RAZONAMIENTO BASADO EN CASOS PARA LA ENSEÑANZA DE TEMAS MÉDICOS

    Directory of Open Access Journals (Sweden)

    Vivian Estrada Sentí

    2002-05-01

    Full Text Available

    Dentro de las técnicas de inteligencia artificial, el Razonamiento Basado en Casos (RBC ocupa un importante lugar ya que facilita el uso de la experiencia acumulada para la toma de decisiones sobre las nuevas situaciones que se presenten. Los autores afirman que el RBC es una importante herramienta de apoyo a la enseñanza de temas médicos, basándose en los resultados de las investigaciones realizadas, y exponen una de las bases de casos utilizadas con estos fines, por su eficiencia para el diagnostico, el pronóstico y la conducta médica.

    Abstract

    In particular, the introduction of artificial intelligence methods and its application in medicine constitute an interesting and perspective scientific field. Specially, Case Based Reasoning (CBR plays an important role in medical science. CBR methods and techniques have been used with success in complex medical decision making. Paper's authors argued CBR represents a very important tool in medical learning and teaching.

  5. Herramienta para programar un controlador lógico programable basado en hardware reconfigurable

    Directory of Open Access Journals (Sweden)

    Valery Moreno Vega

    2011-09-01

    Full Text Available Normal 0 21 false false false MicrosoftInternetExplorer4 /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Tabla normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:10.0pt; font-family:"Times New Roman"; mso-ansi-language:#0400; mso-fareast-language:#0400; mso-bidi-language:#0400;} El presente trabajo se basa en el desarrollo de una herramienta que permita programar un controlador lógico programable (PLC basado en hardware reconfigurable. El desarrollo de la aplicación está  sustentado por una de las metodologías ágiles descrita por XP, utilizando las técnicas de modelación establecidas por el Lenguaje Unificado de Modelado (UML y haciendo uso de una potente herramienta de programación y diseño de interfaces gráficas como es Qt. La programación del PLC se realiza mediante el lenguaje de escalera, atendiendo la norma IEC 61131-3.

  6. Estudio multinivel basado em PISA 2009: determinantes del rendimiento educativo em Uruguay

    Directory of Open Access Journals (Sweden)

    Vivian Tatiane Rodrigues Yuane

    2015-12-01

    Full Text Available Los resultados de PISA 2009 manifiestan que el nivel educativo de los estudiantes Uruguayos se sitúa en una posición por debajo del promedio de los países de la OCDE, además, el promedio de los mismos discrepa entre las regiones del país. Frente a eso, este ensayo se propone a analizar los determinantes del desempeño estudiantil basado en los factores que tradicionalmente la literatura ha identificado como determinantes de la eficacia escolar. Dichos factores están relacionados con el propio alumno, con el centro educativo al que pertenece, su familia y su situación socioeconómica. Este ensayo además, procura controlar los aspectos regionales, para ello, la metodología empleada en este estudio es la regresión multinivel con la que es posible considerar la estructura jerárquica de las variables, el modelo incluye tres niveles: alumno, escuelas y regiones. Los resultados muestran que el nivel educativo de los padres, la calificación de los profesores y el nivel socioeconómico de los alumnos que concurren a la misma escuela, cuanto mayor sean sus valores correspondientes, mayor es el resultado esperado en el rendimiento de los alumnos en el área de comprensión lectora.

  7. Predictor Basado en Prototipos Difusos y Clasificación No-supervisada

    Directory of Open Access Journals (Sweden)

    Aníbal Vásquez

    2015-07-01

    Full Text Available La construcción de prototipos difusos es un método que permite describir a los elementos más representativos de un clúster, a través de su tipicidad. Los prototipos, como los datos más representativos de cada clúster, pueden ser usados en un proceso de clasificación como datos de entrenamiento. Estos prototipos y los clusters pueden ser construidos mediante algoritmos de clustering difuso; los clusters representados por los prototipos poseen variables descriptivas y atributos que pueden ser asociados a nuevos datos. El siguiente trabajo propone una arquitectura que utiliza herramientas de clustering y prototipado difuso, para clasificación no-supervisada y predicción a través de la extracción de variables descriptivas. El desarrollo de un caso de estudio permitió validar el modelo de clasificación para predecir el riesgo de falla en el rendimiento académico de estudiantes, basado en su carga semestral y rendimiento académico, en la selección de cursos antes de registrarse, con un porcentaje de certeza significativo.

  8. Codificador RS(n,k basado en LFCS: caso de estudio RS(7,3

    Directory of Open Access Journals (Sweden)

    Cecilia Sandoval-Ruiz

    2012-01-01

    Full Text Available El presente artículo presenta el diseño de un codi® cador Reed Solomon basado en un circuito concurrente, LFCS - Linear Feedback Concurrent Structure- que permite la generación de los símbolos de redundancia del código de forma paralela, siempre que se le suministren los k símbolos de información a codificar de forma simultánea, el codifi cador ofrece a su salida los símbolos de redundancia correspondientes. Para lograr este desarrollo se generalizó el modelo matemáticos para la descripción del comportamiento del codificador, se realizó la configuración en lenguaje descriptor de hardware VHDL de un codificador Reed Solomon, tomando como caso de estudio el RS(7,3, se simuló el diseño propuesto validando así su funcionamiento, para finalmente realizar la comparación de la implementación del codifi cador entre la versión secuencial y la versión basada en LFCS, obteniendo una reducción de componentes hardware y optimizando la velocidad de respuesta y consumo de potencia. Concluyendo, que el diseño del codi® cador propuesto valida el modelo concurrente generalizado a partir de la correspondencia con la arquitectura del LFCS.

  9. COROTIPOS PRELIMINARES DE PERÚ BASADOS EN LA DISTRIBUCIÓN DE LA FAMILIA ASTERACEAE

    Directory of Open Access Journals (Sweden)

    Berni Britto

    2014-01-01

    Full Text Available El presente estudio se basa en el concepto de categorías corológicas o corotipos para formular una nueva hipótesis de clasificación biogeográfica del Perú basado en la distribución de la familia Astera - ceae. La información recabada dio como resultado que existen 1669 especies de Asteraceae registradas en Perú (hasta el año 2008, distribuidas en 255 géneros. El territorio peruano se dividió en 218 Distritos de Reporte que representan a los 24 departamentos divididos en franjas de 500 m de altitud. La base de datos biogeográficos de Asteraceae se obtuvo cruzando los registros de distribución actualizados con los Distritos de Reporte, expresándose en una matriz de presencia-ausencia. El análisis de datos dio como resultado un total de 14 corotipos preliminares para el Perú: Abiseo, Amotape, Andino, Apurímac-Huan - cavelica, Chachapoyas-Huánuco, Huancabamba, Huascarán, Ica, Lima-Piura, Loreto-Ucayali, Manu, Pasco, Sandia y Tacna.

  10. Las TIC: propuesta para el aprendizaje de enfermería basado en problemas

    Directory of Open Access Journals (Sweden)

    Óscar Boude-Figueredo

    2008-01-01

    Full Text Available Este artículo presenta el análisis de un proceso de investigación cuyo objetivo fue identificar las competencias sobre redes de computadores que alcanzan los estudiantes de enfermería a través del trabajo independiente, en un ambiente de aprendizaje (AA que hace uso de un material educativo digital basado en problemas. En el estudio participaron 22 estudiantes de enfermería de la Universidad de La Sabana, que cursaron la materia telemática durante el segundo semestre de 2007. Se recurrió al estudio de caso, ya que éste permite ver un AA desde todas sus aristas, así como comprender las prácticas y los imaginarios de los actores que intervienen, sus relaciones, tensiones y transformaciones. El 27% de los estudiantes superaron los niveles esperados en el desarrollo de las competencias planteadas, el 63% alcanzó los niveles esperados, y el 14% sólo llegó a los niveles mínimos. Estos logros estuvieron relacionados con las metodologías desarrolladas por cada pareja para solucionar los casos, y con factores tales como el intercambio de saberes con pares, los esquemas de pensamiento propios, la disposición y la actitud de los estudiantes.

  11. Experiencia docente mediante la Metodología de Aprendizaje Basado en Problemas

    Directory of Open Access Journals (Sweden)

    José Manuel López-Guede

    2015-07-01

    Full Text Available En este artículo se describe una experiencia docente de implantación de Aprendizaje Basado en Problemas (ABP llevada a cabo en la Escuela Universitaria de Ingeniería de Vitoria-Gasteiz, de la Universidad del País Vasco (UPV/EHU. La asignatura objeto de la implantación ha sido Arquitectura de Computadores, del Grado en Ingeniería Informática de Gestión y Sistemas de Información. En el artículo se recogen los elementos principales a considerar en toda implantación como son las competencias específicas y transversales a adquirir por el alumnado, por lo que se recoge el temario y los resultados del aprendizaje a lograr y evaluar. Se detalla el problema estructurante diseñado y su división en subproblemas, así como los logros alcanzados, siendo estos altamente satisfactorios tanto para el alumnado en general como para el profesorado.

  12. EL LENGUAJE ORDINARIO: LA CLAVE PARA EL APRENDIZAJE DE LAS MATEMÁTICAS BASADO EN PROBLEMAS

    Directory of Open Access Journals (Sweden)

    José Ángel García Retana

    2015-01-01

    Full Text Available En el año 2012 el Ministerio de Educación Pública de Costa Rica, planteó una nueva propuesta de educación matemática para responder a las exigencias sociales y económicas actuales. Esta propuesta se fundamenta en el aprendizaje basado en problemas (ABP como estrategia metodológica. En el caso del aprendizaje de las matemáticas, tal propuesta demanda considerar la relación que existe entre el lenguaje ordinario y el lenguaje matemático, por cuanto el primero es central en el proceso educativo. Este tipo de aprendizaje se debe conceptualizar en su doble función de herramienta, es decir, para resolver problemas, y como disciplina, dado que el lenguaje matemático permite representar los conceptos que trata, al menos de dos maneras diferentes, la semántica y la gráfico-visual, los cuales en gran medida son determinados por el lenguaje ordinario. Así, el lenguaje ordinario y su campo semántico constituyen el eje transversal para el aprendizaje de esta estrategia metodológica.

  13. El enfoque educativo basado en competencias, un reto que enfrenta la Universidad Veracruzana

    Directory of Open Access Journals (Sweden)

    Adoración Barrales Villegas

    2012-12-01

    Full Text Available Este artículo tiene como propósito compartir la experiencia que ha tenido la Universidad Veracruzana en la promoción e implementación del enfoque educativo basado en competencias. La incorporación obedece a la demanda internacional de formar profesionistas competentes en los ámbitos profesional, personal y social.Una de las acciones relevantes ha sido la capacitación de su personal docente y la implementación de una estrategia denominada Proyecto Aula. Este proyecto va dirigido a sistematizar la práctica docente a partir de la elaboración, aplicación y evaluación de un diseño instruccional para cada una de las experiencias educativas (asignaturas que se imparten al interior de sus programas educativos. Esta es la razón de que se revisen conceptos como competencia, tipos, enfoques, metodología y evaluación por competencias enfatizando los retos y aciertos de nuestra institución.

  14. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

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

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

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

  16. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

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

  17. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

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

  18. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

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

  19. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  20. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  1. Consequence Analysis of the MHTGR and PBMR

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Yang, Joon Eon; Lee, Won Jae

    2006-01-01

    The probabilistic safety assessment of the VHTR design provides a systematic analysis to identify and quantify all risks that the plant imposes to the operators, general public, and the environment and thus demonstrates compliance to regulatory risk criteria. During the preliminary conceptual design of VHTR in Korea, both block- and pebble type-fuel are considered. Therefore, the consequence analysis of VHTR using both types of fuel were made in order to obtain the basic insights for the classification of events and the formation of the PSA framework of the VHTR

  2. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-10

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

  3. Nuclear reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  4. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

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

  5. Desempeño estudiantil con el aprendizaje basado en problemas: habilidades y dificultades

    Directory of Open Access Journals (Sweden)

    Bertha Alicia Olmedo-Buenrostro

    Full Text Available El Aprendizaje Basado en Problemas (ABP es una metodología docente centrada en el estudiante siendo este el protagonista de su propio aprendizaje, facilita la adquisición de conocimientos y ayuda al desarrollo de competencias profesionales específicas y genéricas. El presente trabajo pretende compartir la experiencia que se ha tenido en las Facultades de Medicina y Enfermería de la Universidad de Colima, enfatizando las habilidades y competencias que requieren. Los atributos necesarios para integrar las competencias están organizadas en tres rubros: el saber hacer, el saber teórico y el saber ser. El ABP es una metodología de enseñanza-aprendizaje muy enriquecedora que puede potencializar las capacidades de los alumnos, sin embargo, requiere compromiso por parte de ellos, aprender a desaprender los vicios antes adquiridos, madurez de los alumnos de tal forma que aprendan a manejar adecuadamente sus tiempos. Es fundamental mencionar el papel tan relevante que juega el tutor académico para que el alumno potencialice sus habilidades y destrezas. La falta de capacitación como tutor académico de los profesores limita el desarrollo de los alumnos. Si bien es cierto que el modelo favorece el desarrollo integral del estudiante, es importante considerar que el esquema mental sobre el proceso de enseñanza aprendizaje, las experiencias previas, así como las habilidades adquiridas en los niveles anteriores impactará en el éxito del estudiante en su proceso de formación profesional, sin olvidar el papel importante que juega el tutor para favorecer el aprendizaje en los estudiantes.

  6. Sistema de costos basado en actividades en hoteles cuatro estrellas del estado Mérida, Venezuela

    Directory of Open Access Journals (Sweden)

    Marysela Coromoto Morillo Moreno

    2017-03-01

    Full Text Available Las empresas actualmente demandan sistemas de costos que reporten un mayor detalle en la información generada, con el propósito de orientar la aplicación de estrategias que conduzcan a captar y apropiarse de mayores espacios de mercado, sobre todo cuando la competitividad es elevada; por ello, se formuló un sistema de costos basado en actividades, conocido por sus siglas en inglés como abc (Activity-Based Costing, aplicado a los hoteles de turismo de cuatro estrellas de Mérida, en Venezuela, para el control y reducción de costos en los servicios prestados. A partir de una investigación de carácter exploratorio, con un diseño de campo y con apoyo documental, se halló que el sector hotelero encuentra en el abc una oportunidad para afinar controles sobre los altos y variados costos en los que incurre a partir del conocimiento profundo de actividades consumidas, con sus medidas y frecuencia. De esta manera, se obtiene el costo real de los distintos servicios al vincularlos con las actividades desarrolladas para su obtención; finalmente, se sugiere un conjunto de variables que pudieran fungir como generadoras de valor e indicadoras de eficiencia. Se concluye que el abc representa una herramienta de gestión que orienta las decisiones estratégicas y el control de los costos con la consecuente reducción de los mismos para la maximización de beneficios.

  7. QUÉ SE PUEDE APRENDER DE LA LITERATURA SOBRE EL APRENDIZAJE BASADO EN PROBLEMAS

    Directory of Open Access Journals (Sweden)

    YURI GORBANEFF

    2010-01-01

    Full Text Available El trabajo es una revisión de la literatura sobre la aplicabilidad del aprendizaje basado en problemas (ABP en la enseñanza de la administración. El método de ABP se fundamenta en el constructivismo y teorías derivadas del constructivismo, como la de aprendizaje signifi cativo, las teorías socio cultural, dinámica y cognoscitiva de aprendizaje. El ABP fue aplicado con éxito en la enseñanza de derecho y medicina. Los investigadores que estudian su aplicación en la enseñanza de administración, están de acuerdo en que el ABP contribuye a mejorar la comprensión de lectura, la actitud deliberativa, refl exiva y critica de los alumnos, genera las habilidades de liderazgo, trabajo en equipo, emprendimiento y de aprendizaje autónomo, contribuye a la motivación del alumno por la carrera de administración. Existen los obstáculos para el ABP, la naturaleza del saber administrativo que se apoya sobre la experiencia e intuición, el modesto lugar que la solución de problemas ocupa en la práctica del recién egresado de administración, la ausencia del material didáctico apropiado. Las bondades del ABP ameritan futura investigación para remover los obstáculos que difi cultan su aplicación.

  8. Sistema Inteligente de Supervisión de Alarmas Basado en Microcontroladores PIC, SISAP

    Directory of Open Access Journals (Sweden)

    Ioslán Sánchez Martínez

    2010-09-01

    Full Text Available Normal 0 21 false false false MicrosoftInternetExplorer4 st1:*{behavior:url(#ieooui } /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Tabla normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:10.0pt; font-family:"Times New Roman"; mso-ansi-language:#0400; mso-fareast-language:#0400; mso-bidi-language:#0400;} En este artículo se describe hasta la etapa presente de desarrollo del prototipo SISAP (Sistema Inteligente de Supervisión de Alarmas basado en Microcontroladores PICs desarrollado a partir de una propuesta de la Dirección Territorial de ETECSA de Sancti Spíritus con el fin de incrementar las prestaciones de los sistemas instalados para la supervisión de alarmas tecnológicas en centros no atendidos del territorio.  El dispositivo SISAP se encuentra en la versión de desarrollo 0.5 en estado “no concluido”. Hasta este punto es capaz de manejar hasta 40 eventos, que pueden ser on/off o nivele de voltaje y transmitirlos a través de una interfaz telefónica utilizando un protocolo de tonos DTMF. Palabras Clave: Alarmas, Microcontrolador PIC, Voltajes, Eventos on/off, Tonos DTMF.

  9. Estudio y caracterización de vidriados vitrocerámicos basados en piroxeno

    Directory of Open Access Journals (Sweden)

    Lucas, F.

    2004-10-01

    Full Text Available A method was proposed to develop pyroxene-based glass-ceramic glazes. First, was studied, by X-ray diffraction and scanning electron microscopy, the effect of several additives in the monophasic crystallization of pyroxene from glasses in the CaO•MgO•Al2O3•SiO2 quaternary system. After, it was determined the sinterization intervals and thermal properties of the glasses, by hot stage microscopy and dilatometry. Finally, some studied glasses were chosen and glazed tiles were developed under fast firing wall- and floor-tile industrial cycles. The results proved the reproducibility of the microstructural characteristics obtained in the previous study with glasses. The measurement of different mechanical properties confirmed their potential application in nowadays industrial processing.

    En este trabajo se presenta una metodología para desarrollar vidriados vitrocerámicos basados en piroxeno. En primer lugar se ha estudiado, mediante difracción de rayos X y microscopía electrónica de barrido, el efecto de diferentes aditivos en la cristalización monofásica de piroxeno a partir de vidrios en el sistema CaO•MgO•Al2O3•SiO2. Posteriormente, se han determinado los intervalos de sinterización y características térmicas de los vidrios, mediante microscopía de calefacción y dilatometría. Finalmente, con los vidrios seleccionados se han preparado piezas esmaltadas mediante tratamientos térmicos utilizados habitualmente en la industria de pavimento y revestimiento, comprobando la reproducibilidad de las características microestructurales obtenidas en los estudios iniciales en vidrios. Se han medido sobre las piezas esmaltadas diferentes propiedades que permiten confirmar su potencial aplicación en procesados industriales de monococción y/o bicocción.

  10. Fusión Borrosa de Estimadores para Aplicaciones de Control Basado en Imagen

    Directory of Open Access Journals (Sweden)

    Carlos Perez-Vidal

    2010-04-01

    Full Text Available Resumen: El control visual es una disciplina de gran actualidad dentro del control de robots, y dentro de ésta, los algoritmos de predicción se usan para estimar la localización de objetos o características visuales proporcionadas por un sensor con retardo (cámara. Algunos de los algoritmos más utilizados son: el filtro de Kalman; los filtros alpha-beta/gamma (αβ/γ; el AKF; el SKF; etc. El mayor problema de algunos de ellos es conseguir que su implementación permita trabajar en aplicaciones con fuertes restricciones temporales o de tiempo real. En este artículo se presenta un nuevo método de predicción, denominado FMF, basado en la fusión o combinación borrosa de varios filtros, y por tanto con un alto coste computacional. En el artículo se estudia a través de simulación la mejora obtenida con la predicción del FMF respecto a los filtros individuales, lo que justifica su interés. Así mismo, se desarrolla su implementación de tiempo real en una FPGA empleando técnicas de paralelización y segmentado. La viabilidad, robustez y fiabilidad del algoritmo propuesto se ha comprobado mediante una aplicación experimental de control visual. Palabras clave: Métodos predictivos, algoritmos paralelos, sistemas fuzzy, visión por computador, control automático

  11. MIDDIS: ARQUITECTURA DE REFERENCIA PARA LA INTERACCIÓN DE SERVICIOS BASADOS EN SOA E IMS

    Directory of Open Access Journals (Sweden)

    Ximena Velasco Melo

    2010-01-01

    Full Text Available En telecomunicaciones la tendencia actual está dirigida hacia una búsqueda de la convergencia de redes fijas y móviles, y por lo tanto, las redes que se diseñan son más complejas. Así mismo, se presentan nuevos retos en el campo de la interconexión e integración de servicios a través de múltiples redes, tecnologías y áreas de negocio, lo cual hace imprescindible interoperar los servicios de las Tecnologías de Información (Information Technologies, IT , con los de telecomunicaciones. Para aportar en la solución de estos retos y debido además, a la ausencia de un entorno de telecomunicaciones convergente y completamente adecuado para la prestación de servicios tradicionales y nuevos, en este artículo se presenta una arquitectura de referencia que permite la habilitación y entrega rápida de servicios convergentes para el mundo IT y el mundo de las Telecomunicaciones, con la mediación en la interacción de servicios basados en la Arquitectura Orientada a Servicios (Service Oriented Architecture, SOA, y el Subsistema Multimedia IP (IP Multimedia Subsystem, IMS . La característica esencial del middleware, implementado en un Entorno de Ejecución de Lógica de Servicio (Service Logic Execution Environment, SLEE, consiste en que IMS utiliza a SOA para integrar sus propios elementos software con componentes externos y de esta manera, se logra la combinación de las facilidades de la Web y de IMS para exponer un conjunto de servicios enriquecidos para ambos mundos.

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

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

  13. NEUTRONIC REACTOR

    Science.gov (United States)

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

    1958-04-15

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

  14. Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

    2012-01-01

    Highlights: ► Models for PBMR and thermochemical sulfur cycle based hydrogen plant are developed. ► Models are validated against available data in literature. ► Transient in coupled reactor and hydrogen plant system is studied. ► For loss-of-heat sink accident, temperature feedback within the reactor core enables shut down of the reactor. - Abstract: A high temperature reactor (HTR) is a candidate to drive high temperature water-splitting using process heat. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Three thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. The models and coupling scheme are presented here, as well as a transient test case initiated within the chemical plant. The 50% feed flow failure within the chemical plant results in a slow loss-of-heat sink (LOHS) accident in the nuclear reactor. Due to the temperature feedback within the reactor core the nuclear reactor partially shuts down over 1500 s. Two distinct regions are identified within the coupled plant response: (1) immediate LOHS due to the loss of the sulfuric

  15. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

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

  16. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

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

  17. Reactor facility

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

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

  19. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

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

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

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

  1. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

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

  2. Reactor core

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

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

  4. SEGUROS AGRÍCOLAS BASADOS EN ÍNDICES CLIMÁTICOS: UN ESTUDIO DE CASO EN BOLIVIA

    Directory of Open Access Journals (Sweden)

    Ricardo Nogales Carvajal

    2014-07-01

    Full Text Available Los seguros agrícolas basados en índices climáticos son instrumentos financieros novedosos para la gestión de riesgos de agricultores. A diferencia de los seguros tradicionales, éstos pueden mitigar e incluso anular el riesgo moral y la selección adversa, permitiendo abaratar sus primas. Esta característica hace de este tipo de seguros un mecanismo atractivo para economías en vías de desarrollo con gran parte de población rural sumida en severas condiciones de precariedad y, simultáneamente, se promueve el desarrollo de un mercado de seguros agrícolas. Este tipo de seguros no es aún empleado en Bolivia, pero su promoción por parte del aparato público tiene gran potencial de coadyuvar a la seguridad alimentaria y promover el desarrollo económico. El documento presenta algunas experiencias exitosas en el desarrollo de seguros basados en índices climáticos a nivel mundial y analiza el estado actual de desarrollo de este mercado en Bolivia. Se revisan en detalle los fundamentos técnicos de la creación de este tipo de seguros, a partir de los cuales, se presentan propuestas de esquemas paramétricos de seguros agrícolas basados en índices climáticos para la protección de cultivos de trigo y papa, ambos de ciclo intermedio, en el municipio de Anzaldo (sur-oeste de Cochabamba.

  5. Estado situacional de los modelos basados en agentes y su impacto en la investigación organizacional

    OpenAIRE

    Ovalle Pinilla, Dayana

    2014-01-01

    En un mundo hiperconectado, dinámico y cargado de incertidumbre como el actual, los métodos y modelos analíticos convencionales están mostrando sus limitaciones. Las organizaciones requieren, por tanto, herramientas útiles que empleen tecnología de información y modelos de simulación computacional como mecanismos para la toma de decisiones y la resolución de problemas. Una de las más recientes, potentes y prometedoras es el modelamiento y la simulación basados en agentes (MSBA). Muchas organi...

  6. Aprendizaje basado en problemas: aplicación en la asignatura gestión del conocimiento

    OpenAIRE

    Plaza-Angulo, Juan José; López-Toro, Alberto Antonio

    2017-01-01

    La asignatura Gestión del Conocimiento se imparte en el tercer curso del Grado en Marketing e Investigación de Mercados de la Universidad de Málaga, y forma parte de un proyecto de innovación educativa que pretende aplicar la metodología del aprendizaje basado en problemas (ABP). Esta metodología convierte al alumno en un sujeto activo que debe resolver situaciones reales, comprender su impacto, interpretar datos, diseñar estrategias y utilizar el conocimiento teórico y habilidades que posee ...

  7. Atributos a considerar en la definición de procesos de desarrollo de software basados en modelos de outsourcing

    OpenAIRE

    Trujillo Ballesteros, Yemina Nayhely Haydee

    2005-01-01

    Las empresas dedicadas a la consultaría y desarrollo de software, han diseñado procesos de desarrollo de software distribuido basados en modelos de outsourcing. El outsourcing es la subcontratación de mano de obra de terceros que realiza el trabajo que no está relacionado con el giro de la empresa ** [4]. Los principales estilos de outsourcing son el on-site y off-site con los que se han diseñado distintos modelos y conceptos relacionados con éste. Conocer los modelos de outsourcing y las car...

  8. Modelos flexibles de selección de personal basados en la valoración de competencias.

    Directory of Open Access Journals (Sweden)

    Liern Carrión, Vicente

    2008-01-01

    Full Text Available Las decisiones que toman los directivos respecto a la selección de personalcondicionan fuertemente el éxito de la empresa, pues si los empleados son elegidoscorrectamente suponen una fuente de ventaja competitiva. En este trabajo presentamosalgunos modelos fuzzy de selección de personal basados en la gestión por competencias.y la comparación con las valoraciones que la empresa considera óptimas en cadacompetencia (candidato ideal. Presentamos un algoritmo que permiten establecer unaordenación, incluso cuando el candidato ideal sólo ha sido valorado parcialmente.

  9. ALGORITMO PARA EL APRENDIZAJE DE REGLAS DE CLASIFICACION BASADO EN LA TEORÍA DE LOS CONJUNTOS APROXIMADOS EXTENDIDA

    Directory of Open Access Journals (Sweden)

    YAIMA FILIBERTO

    2011-01-01

    Full Text Available Los conjuntos aproximados han demostrado ser efectivos para desarrollar técnicas de aprendizaje automático, entre ellos métodos para el descubrimiento de reglas de clasificación. En este trabajo se presenta un algoritmo para generar reglas de clasificación basado en relaciones de similaridad, lo que permite que sea aplicable en casos donde los rasgos tienen dominio discreto o continuo. Los resultados experimentales muestran un desempeño satisfactorio en comparación con otros algoritmos conocidos como C4.5 y MODLEM.

  10. Tratamientos basados en la evidencia para adolescentes con trastornos por consumo de cannabis en el Sistema Público de Salud

    OpenAIRE

    Sergio Fernández Artamendi; José Ramón Fernández Hermida; Mark D. Godley; Roberto Secades Villa

    2014-01-01

    El objetivo de este estudio era describir la implementación de dos programas basados en la evidencia (PBE) para adolescentes con trastornos por consumo de cannabis en el Sistema Público de Salud, y sus principales resultados. La Aproximación de Reforzamiento Comunitario para Adolescentes (A-CRA) y el Control de Contingencias (MC) fueron elegidos como los programas de intervención más eficaces para esta población. Un total de 26 adolescentes participaron en el estudio (91,7% chicos; edad media...

  11. Elaboración de un modelo basado en CFD para predecir el comportamiento de un Aerogenerador de Eje Vertical

    OpenAIRE

    Vega Angulo, Carmen Victoria

    2012-01-01

    El presente Trabajo Fin de Máster tiene como objetivo elaborar un modelo numérico basado en la Mecánica de Fluidos Computacional que permita predecir el comportamiento aerodinámico de un aerogenerador de eje vertical. Para este estudio fue seleccionado un aerogenerador tipo H-Darrieus de tres aspas con dimensiones tomadas de la referencia [Gupta, 2010]. Para elaborar el modelo en CFD, fue necesario discretizar el dominio en pequeñas celdas mediante el programa Gambit v.2.3.16 y es...

  12. Desarrollo de un sistema basado en la visión artificial para el reconocimiento de placas vehiculares

    OpenAIRE

    Suárez Pino, José Antonio; Suárez Pino, José Antonio; Suárez Pino, José Antonio

    2011-01-01

    El presente trabajo describe el desarrollo del sistema de reconocimiento de placas vehiculares basado en visión artificial. En primer lugar, la placa es detectada dentro de la imagen digital usando las características de ancho, altura y área de los objetos que hay en la imagen. Después de ubicar la placa, los caracteres son extraídos de la imagen uno a uno. Finalmente, cada carácter es reconocido usando el algoritmo de redes neuronales artificial. Las pruebas de reconocimiento se realizaron t...

  13. Un enfoque basado en competencias de sistemas multiagentes para la enseñanza de Inteligencia Artificial

    OpenAIRE

    Cecchi, Laura; Vaucheret, Claudio A.; Moya, Mario; Kogan, Pablo; Castillo, Rodolfo del; Torres, Guillermo

    2009-01-01

    En este trabajo se presenta un enfoque basado en competencias de sistemas multiagentes como fundamento de la enseñanza de tópicos avanzados en Inteligencia Artificial. La metodología fue implementada en la materia optativa Robótica Cognitiva con alumnos del 5to. año de la carrera Licenciatura en Ciencias de la Computación, de la Universidad Nacional del Comahue, en un dominio que es conocido y divertido: el fútbol. Los campeonatos que se realizan entre los diferentes equipos permiten que los ...

  14. Dispositivo de navegación para personas invidentes basado en la tecnología time of flight

    OpenAIRE

    LENGUA, ISMAEL; DUNAI, LARISA; PERIS FAJARNÉS, GUILLERMO; DEFEZ, BEATRIZ

    2013-01-01

    El artículo presenta un nuevo dispositivo de navegación y detección de obstáculos para las personas ciegas, basado en la tecnología Time-of-Flight y en sonidos acústicos. El dispositivo se ha desarrollado como un dispositivo de ayuda, complementario al bastón, para las personas invidentes. Su objetivo primordial es detectar los obstáculos e informar al usuario mediante sonidos acústicos de la locación de los mismos, tanto en distancia como en dirección. El dispositivo tiene un rango de trabaj...

  15. ENSEÑAR BIOÉTICA A ESTUDIANTES DE MEDICINA MEDIANTE EL APRENDIZAJE BASADO EN PROBLEMAS (ABP)

    OpenAIRE

    JOAQUIM BOSCH-BARRERA; HUGO C. BRICEÑO GARCÍA; DOLORS CAPELLÀ; CARMEN DE CASTRO VILA; RAMON FARRÉS; ANNA QUINTANAS; JOSEP RAMIS; ROSA ROCA; JOAN BRUNET

    2015-01-01

    Se describe la implantación de la asignatura de Bioética en una Facultad de Medicina con el objetivo de dotar a los 73 alumnos de quinto curso de competencias para manejar conflictos éticos en su práctica profesional. El método docente utilizado principalmente fue el aprendizaje basado en problemas. Se describen las competencias y objetivos docentes marcados. El diseño de la asignatura consistió en un seminario teórico (2 horas), un taller práctico (2 horas), cuatro casos de aprendizaje basad...

  16. Validez de un modelo basado en los costes de transacción para identificar los beneficios de los SIIO

    OpenAIRE

    Maggiolini,Piercarlo; Vallès,Ramon Salvador

    2002-01-01

    El objetivo principal de este trabajo de investigación es verificar la validez de un modelo basado en los costes de transacción, para evaluar el impacto de la introducción y uso de los Sistemas de Información Interorganizativos (SIIO) en las empresas. Se propone un modelo que considera diferentes tipos de beneficios, y después, a modo exploratorio, se aplica a la identificación de los beneficios obtenidos por el uso de Sistemas de Intercambio Electrónico de Datos (EDI) en un grupo de empresas...

  17. Plataforma web y sistema integral de incentivos basado en el esquema de cupones online y el mercado de compras colectivas

    OpenAIRE

    Gómez Giraldo, Diego Alejandro

    2013-01-01

    En este trabajo se realiza una descripción y análisis del esquema de funcionamiento detrás del emergente mercado de las compras colectivas y los cupones online desde una perspectiva tanto teórica como empírica. Inicialmente, se desarrolla un marco teórico teniendo en cuenta elementos de: teoría económica, e-marketing y comercio electrónico en los que se basa éste mercado. Posteriormente, se muestra el proyecto de implementación de una plataforma virtual y un sistema de incentivos basado en ...

  18. Gráficos de control basados en la variación de un rango multivariante

    OpenAIRE

    Canales Florencio, Carolina

    2016-01-01

    El presente TFG consiste en implementar una serie de gráficos de control para observaciones multivariantes a partir de un rango basado en la centralidad de los datos. Este tipo de gráficos de control se inspiran en los que introdujo Liu [8], en los cuales para cada observación multivariante se obtiene su rango, es decir la proporción de observaciones históricas que son menos profundas que ella. Para una muestra, de carácter multivariante, y un conjunto de observaciones históricas se obt...

  19. The Design of High Reliability Magnetic Bearing Systems for Helium Cooled Reactor Machinery

    International Nuclear Information System (INIS)

    Swann, M.; Davies, N.; Jayawant, R.; Leung, R.; Shultz, R.; Gao, R.; Guo, Z.

    2014-01-01

    The requirements for magnetic bearing equipped machinery used in high temperature, helium cooled, graphite moderated reactor applications present a set of design considerations that are unlike most other applications of magnetic bearing technology in large industrial rotating equipment, for example as used in the oil and gas or other power generation applications. In particular, the bearings are typically immersed directly in the process gas in order to take advantage of the design simplicity that comes about from the elimination of ancillary lubrication and cooling systems for bearings and seals. Such duty means that the bearings will usually see high temperatures and pressures in service and will also typically be subject to graphite particulate and attendant radioactive contamination over time. In addition, unlike most industrial applications, seismic loading events become of paramount importance for the magnetic bearings system, both for actuators and controls. The auxiliary bearing design requirements, in particular, become especially demanding when one considers that the whole mechanical structure of the magnetic bearing system is located inside an inaccessible pressure vessel that should be rarely, if ever, disassembled over the service life of the power plant. Lastly, many machinery designs for gas cooled nuclear power plants utilize vertical orientation. This circumstance presents its own unique requirements for the machinery dynamics and bearing loads. Based on the authors’ experience with machine design and supply on several helium cooled reactor projects including Ft. St. Vrain (US), GT-MHR (Russia), PBMR (South Africa), GTHTR (Japan), and most recently HTR-PM (China), this paper addresses many of the design considerations for such machinery and how the application of magnetic bearings directly affects machinery reliability and availability, operability, and maintainability. Remote inspection and diagnostics are a key focus of this paper. (author)

  20. Localization of the Hot Spot in the Gap of Pebble Bed of Very High Temperature Gas Cooled Reactor(VHTGR)

    International Nuclear Information System (INIS)

    Lee, Sa Ya; Hong, Sung Je; Lee, Jae Young

    2010-01-01

    Pebble Bed Reactor(PBR) has been investigated intensively due to its benefits in management, but its complicated flow geometry requests reliable analytical methods. Hassan and Lee et al. have been made three dimensional computational methods. Hassan also measured local velocity fields with Particle Tracking Velocimetry(PTV), in small sized packed bed using liquid coolant, and Lee et al. measured flow field in the 2-dimensional wind tunnel with a hot wire system. In the present study, we develop the scaled up wind tunnel of pebble bed to use air as coolant in the same Reynolds number condition, as 21614, of the PBMR-250MWth. In order to measure the local surface temperature, the heating system and temperature measurement system were installed and heat transfer analogy was performed. The local surface temperature data shows that the predicted hot spots by Lee et al. at the top and bottom of the pebble by the velocity field measurement are reasonable, but the heat conduction is prior than contact effect at contact points

  1. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. El aprendizaje basado en problemas: De herejía artificial a res popularis

    Directory of Open Access Journals (Sweden)

    L.A. Branda

    Full Text Available La extensa implementación del aprendizaje basado en problemas (ABP en el proceso enseñanza-aprendizaje ha resultado en su transformación de herejía artificial a res popularis con la consecuente proliferación de publicaciones, libros y congresos sobre el tema. A menudo, esta avalancha de información, ha creado una confusión en la comprensión de qué es el ABP como estrategia de aprendizaje. Este artículo presenta al lector una definición de lo que se consideró que era el ABP y su extensión, además de incluir la resolución de problemas. Se indica la importancia de los objetivos de aprendizaje (resultados del aprendizaje y se presentan algunos pasos que se deben seguir en la preparación de situaciones/escenarios/problemas/casos. De forma general, se describen la evaluación de los estudiantes fundamentalmente formativa, basada en las observaciones hechas en las sesiones de tutoría, y la evaluación de carácter sumativo. La descripción de las etapas más comunes en el ABP tiene el propósito de indicar lo que los estudiantes pueden hacer y no que deben hacer. Si se consideran las limitaciones de recursos que tienen la mayoría de las instituciones que desean implementar el ABP, se describe la aplicación de esta estrategia en grupos grandes. Se discute el rol del tutor facilitador y se indican las características de sus intervenciones en un continuo que va desde jerárquica a facilitadora de la autonomía del estudiante en su aprendizaje. Este artículo finaliza con una reflexión sobre el aprendizaje autodirigido y su relación con el aprendizaje autorregulado.

  3. Aprendizaje basado en problemas: Metodología de aprendizaje centrada en el estudiante

    Directory of Open Access Journals (Sweden)

    V. Tórtora

    2013-11-01

    Full Text Available 8vas Jornadas de la Sociedad de Bioquímica y Biología Molecular. 12 y 13 de setiembre 2013. SIMPOSIO VI: “Educación en Ciencias” Coordinadores: María Noel Álvarez, Andrea MedeirosLa reforma del plan de estudios de la carrera de Doctor en Medicina implicó, entre outras modificaciones, la incorporación de nuevas metodologías de enseñanza y aprendizaje. Una de ellas es la de “aprendizaje basado en problemas” (ABP, que se caracteriza por colocar al estudiante en el centro del proceso de aprendizaje, promoviendo que este sea significativo, además de desarrollar habilidades y competencias indispensables en el ejercicio profesional. El ABP se realiza en grupos de aproximadamente 20 estudiantes, en dos instancias. El problema es diseñado por un equipo docente interdisciplinario, en función a los objetivos del curso. En la primera instancia se presenta el problema y los estudiantes discuten en función de sus conocimientos previos, identificando las áreas de saber y de no saber sobre el problema, e intentan plantear hipótesis explicativas sobre las dificultades identificadas. Lugo deben determinar qué competencias y nuevos conocimientos necesitan para comprobarlas. Este proceso es realizado de manera autónoma por los estudiantes, bajo la dirección de un docente tutor que actúa como facilitador del aprendizaje. Entre una instancia y otra buscan la información necesaria para lograr los objetivos de estudio definidos por el grupo, y regresan al problema en la segunda instancia. La información encontrada es compartida entre los integrantes del grupo, mediante discusión en foros en la plataforma de aprendizaje virtual (EVA y durante las clases presenciales, haciendo que el aprendizaje sea además cooperativo. En esta presentación se expondrá esta metodología de trabajo junto con un ejemplo de un problema específico con objetivos de bioquímica.En esta presentación se expondrá esta metodología de trabajo junto con un ejemplo

  4. TRANSFORMACIÓN DE ESQUEMAS RELACIONALES ORIENTADOS A OBJETOS: UN ENFOQUE BASADO EN EL OBJETO

    Directory of Open Access Journals (Sweden)

    Roberto Sepúlveda Lima

    2007-05-01

    Full Text Available

     

    Las innovaciones en la industria de las bases de datos son permanentes. Entre las más recientes , se tienen las bases de datos distribuidas, el paradigma orientado a objetos (OO, y la tecnología XML*. Todas tienen un objetivo común, mejorar la calidad de los servicios. Por otro lado se observa que la inmensa mayoría de los sistemas de bases de datos (BD actuales tienen una estructura relacional y aún siguen almacenando voluminosas informaciones de sumo valor para las organizaciones. La transformación de los esquemas de bases de datos permanecen siendo un campo de investigación de primordial importancia, ya que se observan numerosas aplicaciones sometidas hoy en día a procesos de reingeniería. Aunque la industria de software ya ofrece algunas herramientas para automatizar este proceso, aún queda mucho por hacer. Y en este contexto, la comprensión del formalismo matemático y de la orientación de los enfoques propuestos no solo podría permitir una mejor comprensión de estos, sino también apoyar a la realización de trabajos futuros en este campo. El presente artículo describe un enfoque de transformación basado en el objeto y en cómo refinarlo mediante traducción de consultas SQL en código C++.

  5. Aprendizaje basado en proyectos aplicados al entorno del laboratorio de inmunología

    Directory of Open Access Journals (Sweden)

    Javier Jiménez Jiménez

    2013-11-01

    Full Text Available En la aplicación de metodologías participativas en el entorno de laboratorio, son muchos los recursos educativos que han mostrado su utilidad para la adquisición de competencias, tanto genéricas (trabajo colaborativo, autoaprendizaje como específicas (razonamiento clínico, reconocimiento de mecanismos relacionados con fundamentos científicos. En este contexto, en la asignatura “Estructura y función de la sangre y del sistema inmunitario” de segundo curso del Grado en Medicina de la UIC se ha introducido una estrategia de aprendizaje basada en proyectos (ABP, que ha permitido dar sentido a la adquisición de conocimientos y habilidades con el fin específico de solucionar un caso clínico. Para ello, los estudiantes, en grupos de tres, se enfrentan al caso debiendo analizar, orientar y seleccionar una prueba de laboratorio que permita confirmar un diagnóstico. El alumno dispone de conexión a internet para la obtención de la base teórica del caso planteado y el apoyo de un tutor para orientar su investigación. Una vez decidida la prueba a realizar dispone de la muestra biológica del paciente, de los kits diagnósticos apropiados y de los protocolos necesarios para llevarlos a cabo. El alumno, con ayuda del tutor, realiza el test diagnóstico que ha seleccionado, obtiene los resultados, los interpreta y con la información obtenida emite un diagnóstico del caso clínico propuesto. Posteriormente los grupos de alumnos exponen cada caso clínico al resto de sus compañeros. En la presente comunicación se muestra con más detalle la realización de este tipo de “Aprendizaje Basado en Proyecto”, y se discute la aceptación que ha tenido por parte de los alumnos, los resultados académicos obtenidos y el grado de adquisición de las competencias trabajadas.

  6. Perfiles de personas con deficiencia mental basados en las funciones motrices gruesas

    Directory of Open Access Journals (Sweden)

    Isabel María FERRÁNDIZ VINDEL

    2009-11-01

    Full Text Available RESUMEN: El propósito de este estudio ha sido identificar posibles subtipos de alumnos con Deficiencia Mental basados en funciones motrices gmesas. Se seleccionó un grupo de alumnos, de un centro específico de Deficiencia Mental, que no presentaron deficiencias motóricas asociadas y se les administraron los subtests de motricidad gruesa del «Test Bruininks-Ozeretsky para medir la eficacia motriz». Los cuatro subtipos resultantes mostraron diferentes perfiles de desarrollo motor. A partir de estos resultados, se recomienda diseñar programas específicos que intenten paliar los trastornos motrices gruesos, analizados para cada uno de los subtipos hallados, y realizar adaptaciones curriculares pertinentes, teniendo en cuenta los perfiles motrices estudiados.ABSTRACT: The goal of this work was to identify some different cathegories of mental retarded people, depending on their gross motor functioning. A group of mental retarded people was selected from a special education school. None of theese students presented any associated motor deficiency. The gross motor subtests from the Bruininks-Ozeretsky test were administered in order to evaluate the motor eficiency of these mental retarded people. The resulting four cathegories showed different motor development profiles. From theese results it's recommended to design curricula-based intended to amiliorate the specific gross motor deficits of each group and making didactical schedules taking into account the motor profiles studied.RESUME: Le propos de cette étude est identifier possibles subtypes d'élèves avec déficience mental basses en fonctions motrices grosses. On a selectioné un groupe d'élèves dans un centre especifique d'insuffisance mental qui ne présentent pas insuffisance motoriques associés, on les a passé les subtests de motricité grosse du test Bruininks-Ozeretsky pour mesurer l'efficacité motrice. Les quattre subtypes resultants montrent des différents profils du

  7. Control Basado en Eventos de Sistemas de Primer Orden Con Retardo

    Directory of Open Access Journals (Sweden)

    Ángel Ruiz

    2013-07-01

    Full Text Available Resumen: La teoría de control PID en su vertiente discreta, apoyándose en una gestión periódica de los muestreos (eventos planificados en tiempo se considera un área madura dentro del paradigma del control automático. Por el contrario, la planificación por eventos deriva, casi inevitablemente, en muestreos asíncronos planteando, problemas adicionales que necesitan ser caracterizados y estudiados. Bajo este escenario, aspectos como la sintonía de los controladores y las condiciones para la estabilidad global o la ausencia de ciclos límite siguen siendo temas que todavía están lejos de ser completamente resueltos. Con el trabajo actual se presenta un nuevo esquema de muestreo y control basado en eventos para sistemas de primer orden con retardo para el que se han analizado los aspectos anteriores. El esquema se basa en el Predictor de Smith para la compensación de los retardos, y en el algoritmo de muestreo SSOD (Symmetric Send- On-Delta para la generación de los eventos. En base a este esquema, se desarrolla el análisis de estabilidad y se propone una metodología de sintonía con una interpretación intuitiva y eficaz. Abstract: PID control theory based on periodic managing of samples has become a well-known area in automatic control. Asynchronous sampling inherent to event-based scheduling causes non-linear dynamics. Under this situation, complex problems arise that must be studied. Issues such as controller tuning, conditions for global stability and the absence of limit cycles are topics that are far from being fully solved yet. In this work, a new event-based scheme of sampling and control for FOPTD processes is presented. The scheme is based on the Smith Predictor structure for delay compensation and the SSOD (Symmetric-Send-On- Delta scheme for events generation. By means of the proposed scheme, a stability analysis is addressed and a simple tuning methodology with effective interpretation is proposed. Palabras clave

  8. Modelo del Costo Basado en la Actividad aplicado a consultas por trazadores de enfermedades cardiovasculares

    Directory of Open Access Journals (Sweden)

    Marteau Silvia A.

    2001-01-01

    Full Text Available OBJETIVO: Hacer un análisis de costos, de la atención médica en consultas externas, mediante la metodología del Costo Basado en la Actividad (ABC, por sus siglas en inglés y en relación con eventos trazadores de enfermedades cardiovasculares de origen isquémico en las instituciones del sector público. MATERIAL Y MÉTODOS: El estudio se basó en consultas por enfermedades o eventos trazadores (n=290 y no trazadores (n=1 710, de una muestra de 2 000 consultas de primera vez de un hospital zonal general de agudos (San Roque de Gonnet, de la provincia de Buenos Aires, Argentina, y se realizó de abril a octubre de 1998. El costo se evaluó con la metodología del ABC. RESULTADOS: El mejoramiento de las actividades de atención en el servicio de Clínica Médica conllevaría un ahorro sustancial en los costos indirectos, equivalente a un porcentaje promedio de 7.11 sobre los productos definidos como consultas por hipertensión arterial (HTA, dislipidemia y diabetes. El ahorro total en el costo unitario por producto que se produciría si se eliminaran las actividades mencionadas, estaría en el orden de 11.78% para el producto HTA, de 13.96% para dislipidemia, de 19.05% para diabetes y de 11.45% para las enfermedades no trazadores. Se asignó o se gastó ineficientemente 66.26% de los costos totales indirectos correspondientes al producto dislipidemia y 61.80% de los correspondientes a diabetes. El costo unitario total de las consultas en el servicio de Clínica Médica, según el método tradicional, es de $22.98, valor que en algunos casos está muy por debajo del costo obtenido a partir del método ABC aplicado en este estudio. CONCLUSIONES: Es necesario trabajar en el rediseño del proceso de atención para evaluar las actividades que no agreguen valor al mismo; éstas únicamente generan molestias y demoras al paciente y provocan ineficiencias en el sistema, dado que se asignan recursos a actividades que no optimizan la gestión y, como

  9. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  10. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

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

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

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

  13. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  14. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

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

  15. Reactor container

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  16. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

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

  17. Nuclear reactors

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

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

  19. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

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

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

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

  1. MLR reactor

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

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

  3. Reactor container

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  4. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  5. Modelo de Awareness Basado en Topologías de Interacción para Espacios Virtuales de Trabajo Colaborativo

    Directory of Open Access Journals (Sweden)

    Edwin Alexander Herrera Saavedra

    2014-08-01

    Full Text Available En un sistema de trabajo colaborativo un grupo de usuarios puede realizar actividades combinando sus capacidades y trabajo para conseguir un determinado objetivo, para esto los miembros del grupo tienen que estar al tanto sobre el estado, cambios y las acciones que otros miembros del grupo están realizando, este tipo de información se conoce como “Información Awareness”. Estos procesos de colaboración han demostrado ser complejos de soportar y desarrollar. Dado esto se han creado varios tipos de awareness tratando de responder como soportar particularidades dentro de los sistemas, pero no existe un modelo general basado en la posibles formas grupales de interacción que sirva como guía en la construcción de awareness. En ese contexto esta investigación propone un modelo de awareness basado en interacciones grupales en los CSCW asociando la información de awareness adecuada para cada topología de interacción. Esto permitirá dar un mejor soporte a la información de tipo awareness en aplicaciones colaborativas.

  6. Prototipado rápido de sistemas de procesado de vídeo basados en el VFBC de Xilinx

    Directory of Open Access Journals (Sweden)

    Luis Manuel Garcés Socarrás

    2013-04-01

    Full Text Available El presente trabajo desarrolla módulos hardware para el prototipado rápido de sistemas de procesado de vídeo basados en el controlador de memoria para fotogramas de vídeo (VFBC de Xilinx. Esta implementación permite el almacenamiento de los fotogramas en memoria externa al dispositivo programable, así como su correcto manejo para el diseño de sistemas de procesado espacio-temporales utilizando el flujo de diseño basado en modelos de Xilinx System Generator. Los módulos hardware son los encargados de la configuración y control de las interfaces de escritura y lectura del VFBC, además de la manipulación de las señales de sincronismo de vídeo para la interconexión de periféricos de entrada y salida.El artículo incluye además la descripción de los módulos elaborados así como el análisis de los resultados del empleo de los mismos en el desarrollo de un demostrador de procesado temporal de vídeo utilizando un detector de movimiento simple sobre una placa Spartan-6 SP605 Evaluation Platform.

  7. Introducción a la Diagnosis de Fallos basada en Modelos mediante Aprendizaje basado en Proyectos

    Directory of Open Access Journals (Sweden)

    Ramon Costa Castelló

    2016-04-01

    Full Text Available Resumen: La diagnosis de fallos basada en modelos es hoy en día un campo maduro dentro de la ingeniería de control que empieza a formar parte de los planes de estudios de grado y postgrado. Sin embargo, la falta de buenos materiales pedagógicos dificulta el proceso de enseñanza / aprendizaje. En este trabajo se muestra cómo una metodología de aprendizaje basada en proyectos se ha utilizado en las sesiones de laboratorio del curso de Diagnosis y Control Tolerante a Fallos del Máster en Automática y Robótica de la UPC utilizando un sistema real de tres depótodos. Los métodos de detección de fallos basados en observadores y la utilización de residuos estructurados para el aislamiento de fallos son introducidos a los estudiantes desde un punto de vista práctico, por medio de un conjunto de ejercicios que se proponen para alcanzar un conjunto de objetivos de aprendizaje. Palabras clave: Detección ;Diagnóstico, Residuos, Fallo, Aprendizaje basado en Proyectos.

  8. Reasoning based in cases applied to diagnosis of electric generators; Razonamiento basado en casos aplicado al diagnostico de generadores electricos

    Energy Technology Data Exchange (ETDEWEB)

    De la Torre Vega, H. Octavio; Garcia Tevillo, Arturo; Campuzano Martinez, Roberto [Instituto de Investigaciones Electricas, Temixco, Morelos (Mexico); Lopez Azamar, Ernesto [Comision Federal de Electricidad (Mexico)

    2000-07-01

    The development of a system for the diagnosis of electrical generators that apply techniques of artificial intelligence, is presented, as it is the reasoning based on cases, to support the work of the diagnosis engineer. This system is part of a system called CADIS, dedicated to the diagnosis of electrical generators out of line and reason of previous articles. In this occasion the characteristics of the reasoning module based on experiences (SirBE) are emphasized, indicating how to make a diagnosis using similar cases and how to edit the system base of experience, using the interactive editor of cases. It is included, in addition, a summarized example which represents a case for SirBE and how the system helps to make a diagnosis. [Spanish] Se presenta el desarrollo de un sistema de diagnostico de generadores electricos que aplica tecnicas de inteligencia artificial, como es el razonamiento basado en casos, para apoyar la labor del ingeniero de diagnostico. Este sistema es parte de un sistema denominado CADIS, dedicado al diagnostico de generadores electricos fuera de linea y motivo de articulos anteriores. En esta ocasion se resaltan las caracteristicas del modulo de razonamiento basado en experiencias (SirBE), indicando como realizar un diagnostico utilizando casos similares y como editar la base de experiencia del sistema utilizando el editor interactivo de casos. Se incluye, ademas, un ejemplo resumido de lo que representa un caso para SiRBE y como el sistema ayuda a realizar un diagnostico.

  9. Modelo de gestión para el suministro de materiales e insumos basado en la demanda

    Directory of Open Access Journals (Sweden)

    Isabel Cristina Arango Palacio

    2014-12-01

    Full Text Available Este artículo hace referencia a la gestión de la cadena de abastecimiento, la cual requiere la sincronización de actividades que incluyen el flujo de información e inventarios. Una gestión exitosa depende de la óptima integración y sincronización de las actividades e inventarios requeridos para satisfacer la demanda. A partir de la investigación realizada en una empresa de cosméticos de venta directa, se analizan las estrategias para mejorar la gestión del suministro enfocado a la demanda, de modo que se pueda responder al mercado y, al mismo tiempo, controlar los inventarios y evitar los agotados. En esta investigación se toman en consideración que el suministro basado en el consumo diario, con entregas frecuentes y lotes pequeños flexibiliza el abastecimiento con los cambios de la demanda, y el stock de seguridad en cada empresa de la cadena de abastecimiento se convierte en un amortiguador para proteger la variabilidad. La metodología comienza con la identificación del problema, las posibles causas y las relaciones con los efectos encontrados con el fin de obtener un modelo de gestión del abastecimiento basado en la demanda, para responder rápidamente a la sobredemanda sin generar agotados y, a la vez, reaccionar a la subdemanda sin exceso de inventario.

  10. Diseño de un modelo de negocio para una empresa de desarrollo de software basado en la metodología Lean Startup

    OpenAIRE

    Rodríguez Guerrero, Gianni Alexander

    2015-01-01

    1. Introducción.--2. Planteamiento de la Propuesta de Trabajo.-- 3. Marco Teórico.-- 4. Metodología.-- 5. Resultados.-- 6. Conclusiones y Recomendaciones El presente trabajo de investigación tiene como título ?Diseño de un Modelo de Negocio para una empresa de desarrollo de software basado en la Metodología Lean Startup?, cuya meta fue diseñar un Modelo de Negocio, basado en la metodología Lean Startup, que ayude a empresas de software de la ciudad de Ambato a diseñar proyectos que se ajus...

  11. Materiales composites micro- y nano-estructurados basados en hidróxidos dobles laminares de tipo hidrotalcita y silicatos de la familia de las arcillas.

    OpenAIRE

    Aranda, Pilar; Gómez Avilés, Almudena; Ruiz-Hitzky, Eduardo

    2008-01-01

    Materiales composites micro- y nano-estructurados basados en hidróxidos dobles laminares de tipo hidrotalcita y silicatos de la familia de las arcillas. La presente invención se refiere a materiales composites micro- o nano-estructurados basados en hidróxidos dobles laminares de tipo hidrotalcita y silicatos de la familia de las arcillas. La invención también se refiere al procedimiento de preparación de estos materiales así como a su uso en aplicaciones diversas tales...

  12. Cualidades tecnológicas de uso de la madera de cinco especies forestales, basado en el conocimiento del tejido secundario de la rama, Loreto

    OpenAIRE

    Valderrama Freyre, Heiter; Universidad Nacional de la Amazonía Peruana

    2016-01-01

    Se determinó el comportamiento de la rama de cinco especies forestales del Jardín Botánico Arboretum El Huayo, ubicado en el Centro de Investigación y Enseñanza Forestal (Ciefor) Puerto Almendra en Loreto, Perú, basado en las características anatómicas de la madera de la rama y que se relacionan con la resistencia mecánica, secado, preservado, trabajabilidad, durabilidad natural y fabricación de pulpa para papel. La metodología para determinar el comportamiento tecnológico, basado en el conoc...

  13. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

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

  14. Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico

    International Nuclear Information System (INIS)

    Barragan M, A.M.; Martin del Campo M, C.

    2007-01-01

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

  15. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  16. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  17. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  18. Simulacion borrosa de un reactor con reaccion exotermica no lineal

    Directory of Open Access Journals (Sweden)

    MIGUEL ANGEL RODRIGUEZ BORROTO

    2007-01-01

    Full Text Available En el presente trabajo se desarrolla un modelo difuso basado en la estructura Takagi-Sugeno-Kang dinámica para un reactor continuo de tanque con agitador (RCTA con reacción química de primer orden exotérmico. A partir de datos experimentales obtenidos mediante simulación del proceso real, se obtiene la base de datos de las variables de entrada y salida del proceso y a partir de la misma se elaboran los archivos de datos de entrenamiento y de verificación del modelo borroso el cual es obtenido mediante la herramienta anfis de MATLAB. El modelo obtenido permite predecir la salida del sistema con errores de predicción muy bajos, por lo que el mismo sienta las bases para el diseño de un controlador predictivo no lineal del mismo en próximas etapas de la investigación

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

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

  20. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. Nuclear reactors

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  2. Reactor container

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  3. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

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

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

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

  5. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

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

  6. Hybrid adsorptive membrane reactor

    Science.gov (United States)

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

    2011-03-01

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

  7. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

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

  8. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

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

  9. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

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

  10. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

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

  11. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

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

  12. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

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

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

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

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

  14. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

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

  15. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

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

  16. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

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

  17. Control predictivo económico de vehículos híbridos basados en pilas de combustible

    OpenAIRE

    Sampietro, Jose Luis; Costa Castelló, Ramon; Puig Cayuela, Vicenç

    2015-01-01

    Las pilas de combustible que utilizan el hidrógeno como combustible están siendo consideradas, en estos últimos años, como una alternativa a los combustibles fósiles para su uso en automóviles. Dicha tecnología se puede aplicar en los vehículos de propulsión híbrida. Este trabajo introduce el control predictivo económico (EMPC, siglas en inglés) como técnica de gestión óptima de la energía. Finalmente, se presentan simulaciones de varios escenarios, basados en un control EMPC, en donde se ...

  18. Efecto del ejercicio físico basado en el juego en la leucemia linfocítica aguda

    OpenAIRE

    González García, Laura Ximena; Escobar Zabala, Paola Andrea

    2012-01-01

    Objetivo: Establecer el efecto de un programa de ejercicio físico (EF) basado en el juego sobre el Síndrome de Desacondicionamiento Físico (SDF) de niños con Leucemia Linfocítica Aguda (LLA) entre 5-12 años. Materiales y Métodos: Se realizó un estudio cuasi-experimental con la participación de siete niños tratados por LLA en el Instituto Nacional de Cancerología (INC). Se hizo una evaluación inicial de los determinantes de la condición física (capacidad aeróbica, fuerza muscular, flexibil...

  19. Proceso de Identificación de Comportamiento de Comunidades Educativas Basado en Resultados Académicos

    Directory of Open Access Journals (Sweden)

    Pablo Cigliuti

    2014-12-01

    Full Text Available Los procesos de explotación de información se incorporan al ámbito educativo para ayudar a entender y mejorar tanto la enseñanza de los docentes como el aprendizaje de los alumnos. Entre estas cuestiones se destaca el análisis del comportamiento de comunidades educativas de forma tal de proveer al docente herramientas que ayuden a mejorar la enseñanza/aprendizaje. En este contexto, el presente trabajo tiene como objetivo proponer, estudiar y validar un proceso de explotación de información que permita identificar el comportamiento de comunidades educativas basado en resultados académicos.

  20. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  3. Reactor container

    International Nuclear Information System (INIS)

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

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  5. Un Nuevo Método de Identificación Basado en la Respuesta Escalón en Lazo Abierto de Sistemas Sobre-amortiguados

    Directory of Open Access Journals (Sweden)

    Luis A. Mora

    2017-01-01

    Full Text Available Resumen: En este documento se presenta un nuevo método de identificación basado en la respuesta escalón en lazo abierto de sistemas sobre-amortiguados. Inicialmente el comportamiento transitorio de funciones de transferencia con polos reales múltiples es analizado, estableciendo formulas analíticas para estimar los tiempos que tardan dichos sistemas en alcanzar determinados puntos de su curva de reacción ante estímulos del tipo escalón, incluyendo el tiempo de asentamiento para los criterios del 2% y 5% de error. Posteriormente se ha desarrollado un método de identificación utilizando 3 puntos característicos (tk, yk de la curva de reacción, permitiendo estimar el orden proceso N, la constante de tiempo T de los polos múltiples y el tiempo muerto L del modelo aproximado. Para evaluar el rendimiento del método propuesto, 3 funciones de transferencia de fase mínima y 1 de fase no mínima son utilizadas para comparar con otros métodos de identificación, así como también, un modelo de un reactor químico no isotérmico; estableciendo como índice de eficiencia la raíz del error cuadrático promedio (RMSE de la respuesta transitoria y en frecuencia de los modelos identificados respecto a las respuestas de los sistemas originales con diferentes valores relación señal a ruido (SNR. A partir de los resultados obtenidos se observó que el método propuesto logra buenas aproximaciones de las respuestas transitorias y en frecuencia, siendo un método sencillo y que requiere una baja carga computacional en comparación con otros métodos. Abstract: In this paper a new process identification method based in open loop step response of overdamped systems is presented. Initially the transient behavior of transfer functions with multiple real poles is analyzed, establishing analytical formulas to estimate the time it takes for these systems to reach certain points of his reaction curve to stimuli of the step type, including settling time

  6. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  7. TORT-TD/ATTICA3D: a coupled neutron transport and thermal hydraulics code system for 3-D transient analysis of gas cooled high temperature reactors

    International Nuclear Information System (INIS)

    Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.

    2011-01-01

    Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)

  8. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  9. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  10. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  11. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  12. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

  13. An advanced three-dimensional simulation system for safety analysis of gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lapins, Janis

    2016-07-01

    simulated simultaneously. TORT-TD and ATTICA{sup 3D} exchange data (power distributions or fuel and moderator temperature distribution, possibly hydrogen distribution) by means of a common interface that interpolates values that are exchanged on mutual computational grids by volumetric averaging. As verification for the proper operation of the interface, the steady state of the transient PBMR-400 benchmark was used. After obtaining a coupled steady state, the transient exercises are performed to test the proper working of the interface in time dependent cases. Here, the cold helium ingress, the total control rod withdrawal case and the total control rod ejection case were simulated and compared to results of other partakers of the benchmark. Also, the coupled system was validated for a full power temperature distribution experiment in the Chinese experimental reactor HTR-10 where good agreement could be reached with the measurements. The coupled HTR simulation system TORT-TD/ATTICA{sup 3D} was then applied for single control rod ejection cases for both the PBMR-400 and the HTR-PM. These cases require a 180 model of the reactor. As preparatory works, the control rod cross sections were adjusted to yield the same reactivity increase as the grey curtain model for the PBMR and with MCNP5 for the HTR-PM. Since there are strong shielding effects by neighbouring rods, the power increase was moderate due to strong Doppler and moderator feedbacks. For the HTR-PM, coupled calculations for water ingress cases are simulated. This also tested the whole computational sequence, i.e. steam transport into the core by ATTICA{sup 3D}, then transfer of hydrogen densities (from hydrogen or from steam) to TORT-TD via the interface, interpolation of the macroscopic cross sections which changes the power density, and the feedback to ATTICA{sup 3D}. Additionally, an anticipated transient without scram is simulated where shutdown of the reactor is achieved by the temperature feedback effects. For

  14. Influencia del aprendizaje basado en problemas en la actitud ambiental de los estudiantes de la Institución Educativa “José Carlos Mariátegui” Pampachacra Huancavelica 2014

    OpenAIRE

    Torres Acevedo, Christian Luis

    2017-01-01

    El título de la investigación fue: Influencia del Aprendizaje Basado en Problemas en la Actitud Ambiental de los estudiantes de la Institución Educativa José Carlos Mariátegui Pampachacra - Huancavelica – 2014. El objetivo: Determinar la influencia de la aplicación del aprendizaje basado en problemas en la actitud ambiental de los estudiantes de la Institución Educativa José Carlos Mariátegui Pampachacra - Huancavelica – 2014. La hipótesis: La aplicación del aprendizaje basado en problemas in...

  15. An efficient hybrid sulfur process using PEM electrolysis with a bayonet decomposition reactor - HTR2008-58207

    International Nuclear Information System (INIS)

    Gorensek, M. B.; Summers, W. A.; Lahoda, E. J.; Bolthrunis, C. O.; Greyvenstein, R.

    2008-01-01

    The Hybrid Sulfur (HyS) Process is being developed to produce hydrogen by water-splitting using heat from advanced nuclear reactors. It has the potential for high efficiency and competitive hydrogen production cost, and has been demonstrated at a laboratory scale. As a two-step process, the HyS is one of the simplest thermochemical cycles. The sulfuric acid decomposition reaction is common to all sulfur cycles, including the Sulfur-Iodine (SI) cycle. What distinguishes the HyS Process from the other sulfur cycles is the use of sulfur dioxide (SO 2 ) to depolarize the anode of a water electrolyzer. The two critical HyS Process components are the SO 2 - depolarized electrolyzer (SDE), and the high-temperature decomposition reactor. A proton exchange membrane (PEM)- type SDE and a silicon carbide bayonet-type high-temperature decomposition reactor are being developed for DOE's Nuclear Hydrogen Initiative (NHI) by Savannah River National Laboratory (SRNL) and by Sandia National Laboratories (SNL), respectively. The ultimate goal of the NHI-sponsored work is to couple the SDE and the reactor in an integrated laboratory scale experiment to prove the technical readiness of the HyS cycle for the NGNP demonstration. This paper describes the flowsheet that is being prepared to combine these two components into a viable process and presents the latest performance projections and economics for a HyS Process coupled to a PBMR heat source. The basic flowsheet for this process has been described elsewhere [4]. It requires an acid concentration section because the SDE product, which is limited to no more than 50% H 2 SO 4 by cell voltage considerations, is too dilute to be fed directly to the bayonet, which needs at least 65% H 2 SO 4 in the feed for acceptable performance. Optimization involved trade-offs between decomposition reaction and acid concentration heat requirements. The PBMR heat source can split its heat output between the decomposition reaction and either steam

  16. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

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

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

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

  18. Plutonium and minor actinides management in thermal high - temperature reactors - the EU FP6 project puma

    International Nuclear Information System (INIS)

    Kuijper, J. C.

    2007-01-01

    key elements, which are not covered by these other projects. Earlier projects indicate favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs will be investigated to optimise the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprises the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. It is also envisaged to optimise present Pu CP design and to explore feasibility for MA fuel. New CP designs will be explored that can withstand very high burn-ups and are well adapted for disposal after irradiation. The project benefits greatly from access to past knowledge from Belgonucleaire's Pu HTR fuel irradiation tests of the 1970-s, and also secures access to materials made at that time. (Very) High Temperature Reactor (VHTR) Pu/MA transmuters are envisaged to operate in a global system of various reactor systems and fuel cycle facilities. Fuel cycle studies are envisaged to study the symbiosis to LWR, GCFR and ADS, and to quantify waste streams and radio toxic inventories. The technical, economic, environmental and socio-political impact will be assessed as well. The PUMA project runs from September 1, 2006, until August 31, 2009, and is being executed by a consortium of 15 European partner organisations and one from the USA. The paper presents an overview of planned activities and preliminary/expected results

  19. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  20. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Yasutomi, Yoshiyuki; Nakagawa, Moroo; Sawai, Yuichi; Chiba, Akio; Suzuki, Yasutaka.

    1997-01-01

    Silicon composited with reinforcing metals is used for a divertor cooling substrate having an effect as a cooling tube to provide a silicon base composite material having increased electric resistance and toughness. The blending ratio of reinforcing materials in the form of granules, whiskers or long fibers is controlled in order to control heat conductivity, electric resistivity and mechanical performances. The divertor cooling substrate comprising the silicon base composite material is integrated with a plasma facing material. The production method therefor includes ordinary metal matrix composite forming methods such as powder metallurgy, melting penetration method, high pressure solidification casting method, centrifugal casting method and vacuum casting method. Since the cooling plate is constituted with the light metal and highly electric resistant metal base composite material, sharing force due to eddy current can be reduced, and radiation exposure can be minimized. Accordingly, a cooling structure for a thermonuclear reactor effective for the improvement of environmental problems caused by waste disposal can be attained. (N.H.)

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  2. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  3. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    Energy Technology Data Exchange (ETDEWEB)

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest

  4. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  5. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  7. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  8. Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

    International Nuclear Information System (INIS)

    Brown, N.R.; Revankar, S.T.; Seker, V.; Downar, Th.J.

    2010-01-01

    In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO 3 decomposition section of the SI cycle is proposed and modelled. Since SO 3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear

  9. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  10. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  11. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  12. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  13. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

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

  14. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  16. Efecto de un programa de entrenamiento físico basado en la secuencia de desarrollo sobre el balance postural en futbolistas: ensayo controlado aleatorizado

    OpenAIRE

    Erika Mancera Soto; Édgar Hernández Álvarez; Fabián Hernández Salinas; Laura Prieto Mondragón; Leidy Quiroga Díaz

    2013-01-01

    Antecedentes. En la actividad deportiva, el balance postural es requerido para mantener la estabilidad durante el juego. Por tanto, existe una necesidad de determinar si el aprendizaje motor desde posiciones funcionales, movimientos coordinados están implicados en los deportes. Objetivo. El objetivo de este estudio fue determinar el efecto de un entrenamiento físico basado en la secuencia de desarrollo sobre el balance postural en futbolistas de la selección de la ...

  17. Efecto de un programa de entrenamiento físico basado en la secuencia de desarrollo sobre el balance postural en futbolistas: ensayo controlado aleatorizado

    OpenAIRE

    Mancera Soto, Erika; Hernández Álvarez, Édgar; Hernández Salinas, Fabián; Prieto Mondragón, Laura; Quiroga Díaz, Leidy

    2014-01-01

    Antecedentes. En la actividad deportiva, el balance postural es requerido para mantener la estabilidad durante el juego. Por tanto, existe una necesidad de determinar si el aprendizaje motor desde posiciones funcionales, movimientos coordinados están implicados en los deportes. Objetivo. El objetivo de este estudio fue determinar el efecto de un entrenamiento físico basado en la secuencia de desarrollo sobre el balance postural en futbolistas de la selección de la Universidad Nacional de Colo...

  18. Migración de un SGSI basado en ISO/IEC 27001:2005 a la versión ISO/IEC 27001:2013

    OpenAIRE

    Espol; Rendón Freire, María José

    2015-01-01

    Este proyecto tiene como objetivo identificar los cambios en los requisitos de la norma ISO/IEC 27001:2013 respecto a la versión 2005, para lograr la actualización de un sistema de gestión de seguridad de la información basado en ISO/IEC 27001:2005 acorde a lo establecido en el nuevo estándar. Guayaquil Magíster en Seguridad Informática Aplicada

  19. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  20. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  1. Empleo del aprendizaje basado en problemas (abp. Una propuesta para acercarse a la química verde

    Directory of Open Access Journals (Sweden)

    Marina Lucía Morales Galicia

    2013-05-01

    Full Text Available Los profesores han reflexionado acerca de que la educación ambiental deberá ser formativa, por lo que es importante participar para cambiar esta situación dentro del contexto en la enseñanza de las Ciencias Naturales como parte de incrementar el conocimiento acerca del deterioro del medio ambiente. Este trabajo tiene el propósito de presentar la estrategia Aprendizaje Basado en Problemas (ABP. Para demostrar esta metodología, se presentan los resultados que con esta se obtuvieron gracias a la colaboración tanto de un grupo de alumnos mexicanos del nivel bachillerato, así como de profesores participantes chilenos del nivel secundaria (estudios equivalentes en México. El empleo de esta herramienta didáctica proporcionó un aprendizaje significativo en alumnos y profesores, quienes emplearon sus conocimientos previos para abordar un nuevo aprendizaje, quedando realmente impresionados al enterarse de que el problema propuesto sucedía en su país y de toda la secuela ambiental que generaría el que se siguiera sucediendo o se llevara a cabo. Se consiguió que los participantes aportaran ideas novedosas durante el desarrollo de su trabajo y que aceptaran con gusto la responsabilidad para hacer suyos problemas reales que son parte de su cotidianidad.

  2. RECUPERACIÓN DE IMÁGENES EN LA WEB: SISTEMA PROTOTIPO BASADO EN CONTENIDO Y MANEJO DE CALIDAD

    Directory of Open Access Journals (Sweden)

    Bell Manrique Losada

    2007-01-01

    Full Text Available El presente artículo muestra una revisión de las investigaciones realizadas acerca de la recuperación de imágenes en colecciones digitales en la web, basado en su contenido y en el manejo de la calidad de la información. Teniendo en cuenta este estado del arte, se presenta una propuesta metodológica para el diseño e implementación de un prototipo de sistema de recuperación de imágenes que utilice estas dos técnicas de búsqueda: recuperación basada en contenido y recuperación manejando calidad de la información y, por último, se muestran los avances en el desarrollo de la propuesta.This article presents a compilation of the researches carried on the image recovery from web digital collections, based upon their contents and data quality management. Using this state-of-the-art study as a support, the article presents a methodological proposal to design and implement a system prototype in image recovery that uses these two search techniques; and finally the article presents the progress achieved in developing this proposal.

  3. Enfoque “Aprendizaje Basado en Proyectos” para enseñar sistemas de potencia de gas y vapor

    Directory of Open Access Journals (Sweden)

    Asier Aranzábal

    2014-12-01

    Full Text Available El objetivo de este trabajo es presentar la experiencia y los resultados derivados de la aplicación del método Aprendizaje Basado en Proyectos (PBL en la asignatura Termotecnia de la Titulación de Ingeniería Química de la Universidad del País Vasco/Euskal Herriko Unibertsitatea, para aprender sistemas de potencia de gas y vapor. Tras un análisis crítico de los resultados académicos de los alumnos que aprenden estos sistemas, se observa que la estrategia de enseñanza-aprendizaje tradicional, basada en clases “magistrales + clases de aplicación en una lista de ejercicios (totalmente acotados y con una solución única, estaba fallando. Ante esta situación se plantea un enfoque constructivista centrado en el alumno para que él mismo construya su aprendizaje activa y cooperativamente, y no escuchando clases magistrales, ni memorizando.

  4. Un enfoque basado en competencias de sistemas multiagentes para la enseñanza de Inteligencia Artificial

    Directory of Open Access Journals (Sweden)

    Cecchi, Laura

    2010-01-01

    Full Text Available En este trabajo se presenta un enfoque basado en competencias de sistemas multiagentes como fundamento de la enseñanza de tópicos avanzados en Inteligencia Artificial. La metodología fue implementada en la materia optativa Robótica Cognitiva con alumnos del 5to. año de la carrera Licenciatura en Ciencias de la Computación, de la Universidad Nacional del Comahue, en un dominio que es conocido y divertido: el fútbol. Los campeonatos que se realizan entre los diferentes equipos permiten que los estudiantes evalúen y comparen los resultados. La motivación que se logra es fundamental para despertar interés en el estudio de técnicas la Inteligencia Artificial y en la investigación en general. Se presentan la experiencia desarrollada, un análisis de la metodología y el impacto en el dominio académico y de investigación.

  5. Análisis de experiencias de aprendizajes basados en proyectos: prácticas colaborativas B-Learning

    Directory of Open Access Journals (Sweden)

    Juan Manuel Trujillo Torres

    2015-01-01

    Full Text Available Viene sucediendo de manera continuada cierto distanciamiento entre espacio educativo y realidad social afirmándose esa relación antagónica entre ambos. Bajo esta consideración, el cambio didáctico-pedagógico y organizativo se conforma, hoy más que nunca, como un reclamo que al mismo tiempo ambiciona aprovechar la integración de las TIC para establecer significativos los procesos de aprendizaje y enseñanza y considerar la gestión y practicidad del conocimiento. Así, los aprendizajes basados en problemas (PBL se conforman como una oportunidad para la comunicación, la colaboración, el compromiso, el ejercicio de un sentimiento transformacional, la innovación reflexiva y crítica, entre otras. De este modo, nuestra experiencia b-learning con herramientas web 2.0, pretende convertirse en una oportunidad para la expresión compartida que irradie en forma significativa un pensamiento vivido fruto de la interacción efectiva con el medio desde el ejercicio de la competencia emocional.

  6. JMat - Herramienta remota de cálculo y multiusuario para el aprendizaje basado en problemas usando Matlab

    Directory of Open Access Journals (Sweden)

    Bladimir Bacca Cortes

    2011-01-01

    Full Text Available JMat es una herramienta de cálculo basada en JAVA y EJS (Easy Java Simulations, con un esquema cliente / servidor, soporte multi-usuario y acceso remoto a Matlab. La aplicación está orientada a brindar a los usuarios una interacción con Matlab usando tres interfaces: Consola de Comandos, donde se invocan remotamente comandos de texto compatibles con Matlab. Espacio de Trabajo y Graficación, donde se mantiene un registro automático de las variables de usuario y se grafican individualmente. Funciones de usuario y Transferencia de Archivos, donde el usuario crea sus funciones, envía y recibe datos hacia y desde el servidor. JMat requiere un acceso a Internet, un servidor remoto donde esté instalado Matlab y un cliente (Navegador WEB o aplicación. No se requiere Matlab en el cliente. JMat está siendo usada actualmente en la Universidad del Valle en los cursos de Control Automático de Procesos, Control Inteligente, Redes Neuronales Artificiales, Procesamiento de Señales y Tratamiento Digital de Imágenes como herramienta para el aprendizaje basado en problemas empleando la plataforma de eLearning de la Universidad del Valle.

  7. DESARROLLO DE UN SOFTWARE BASADO EN JAVA CAPAZ DE INTERPRETAR TRAZAS DE ARCHIVOS GENERADOS POR NS-2

    Directory of Open Access Journals (Sweden)

    Fredy Pachón Barbera

    2013-09-01

    Full Text Available En la simulación de una red de datos, se trata de evaluar, investigar y/o analizar su comportamiento  en un ambiente real, siempre buscando la optimización los recursos. El análisis de la información es un punto clave para conocer fortalezas y debilidades de la red simulada. El objetivo de este artículo es mostrar el desarrollo de una herramienta de interpretación de trazas generadas por NS-2 basado en el lenguaje de programación Java  y las variables estadísticas que se tuvieron en cuenta para la implementación, con el fin de facilitar el análisis de redes simuladas con la herramienta network simluator. Finalmente se logra la construcción del software, logrando determinar los cálculos de jitter, throughput y bytes enviados por cada nodo, a partir de las trazas generadas por NS-2.

  8. Programa para la mejora del bienestar de las personas mayores. Estudio piloto basado en la psicología positiva

    Directory of Open Access Journals (Sweden)

    María Guadalupe Jiménez

    2016-01-01

    Full Text Available El objetivo de este trabajo fue probar la eficacia de un programa piloto basado en la psicología positiva y destinado a incrementar el bienestar emocional de las personas mayores. El diseño de investigación fue experimental, con grupos de intervención y control. La muestra estuvo compuesta por 67 adultos de 60 a 89 años de edad. El programa consistió en nueve sesiones de 1.5 horas de duración y frecuencia semanal. Los temas incluyeron fortalezas de carácter, emociones positivas y regulación emocional. Se evaluó el afecto, el nivel de felicidad, el nivel de preocupación, el optimismo y la presión arterial. Los resultados indicaron que los participantes del programa incrementaron significativamente su nivel de felicidad y disminuyeron el nivel de preocupación y la presión arterial sistólica. El incremento del nivel de felicidad en personas mayores favorece la construcción de recursos personales y la implicación con objetivos y proyectos que les acercan al envejecimiento activo y saludable. Los resultados, limitaciones y mejoras de este trabajo son discutidos en el contexto de la psicología positiva y la psicología de la vejez.

  9. El lenguaje ordinario: La clave para el aprendizaje de las matemáticas basados en problemas

    Directory of Open Access Journals (Sweden)

    García Retana, José Ángel

    2015-01-01

    Full Text Available En el año 2012 el Ministerio de Educación Pública de Costa Rica, planteó una nueva propuesta de educación matemática para responder a las exigencias sociales y económicas actuales. Esta propuesta se fundamenta en el aprendizaje basado en problemas (ABP como estrategia metodológica. En el caso del aprendizaje de las matemáticas, tal propuesta demanda considerar la relación que existe entre el lenguaje ordinario y el lenguaje matemático, por cuanto el primero es central en el proceso educativo. Este tipo de aprendizaje se debe conceptualizar en su doble función de herramienta, es decir, para resolver problemas, y como disciplina, dado que el lenguaje matemático permite representar los conceptos que trata, al menos de dos maneras diferentes, la semántica y la gráfico-visual, los cuales en gran medida son determinados por el lenguaje ordinario. Así, el lenguaje ordinario y su campo semántico constituyen el eje transversal para el aprendizaje de esta estrategia metodológica

  10. Desarrollo del pensamiento crítico mediante la aplicación del Aprendizaje Basado en Problemas

    Directory of Open Access Journals (Sweden)

    Verónica Lara Quintero

    Full Text Available Resumen El proceso educativo actual involucra al docente y al estudiante con roles activos aplicando nuevas estrategias. El objetivo del estudio fue determinar si la aplicación del Aprendizaje Basado en Problemas a estudiantes de Ingeniería Biomédica de una universidad privada de Bogotá, favorece la obtención de competencias genéricas, especialmente el pensamiento crítico. El tipo de estudio fue de naturaleza mixta, cuasi-experimental y transaccional. Se utilizó para el enfoque cuantitativo el cuestionario de las Competencias Genéricas Individuales validado por Olivares y Wong (2013 asociado a tres dimensiones del pensamiento crítico: interpretación, juicio e inferência. A su vez, se evaluó de manera cualitativa mediante rúbrica de cuatro categorías relacionadas con la estrategia didáctica: autonomía, participación, comunicación y disposición al pensamiento crítico. Aunque el enfoque cuantitativo no arrojó resultados determinantes en el cambio del pensamiento crítico, si se encontraron cambios a través del análisis cualitativo, especialmente en análisis, interpretación y evaluación.

  11. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  12. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  13. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  14. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  15. Control for nuclear reactor

    International Nuclear Information System (INIS)

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  16. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  17. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  18. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  19. continuous stirred tank reactor (CSTR)

    African Journals Online (AJOL)

    AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... stirred tank reactor (CSTR) and the small and large intestines as plug flow reactor (PFR) ... from the two equations are used for the reactor sizing of the modeled reactors.

  20. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

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

  1. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  2. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  3. Reactor utilization, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1984-01-01

    Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

  4. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  5. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  6. Pagos basados en acciones: concepto, ámbito de ap licación y metodologías de valoración

    Directory of Open Access Journals (Sweden)

    Vernor Mesén Figueroa

    2010-01-01

    Full Text Available La constante evolución de las prácticas empresariales ha hecho que los gerentes y administradores de las empresas, hayan optado por crear novedosas opciones por medio de las cuales puedan interactuar con terceras partes, es así como surgen los pagos basados en acciones, mecanismo que resulta ser una interesante opción por medio de la cual las entidades logran no sólo remunerar de forma atractiva y competitiva a sus empleados y proveedores, sino que también les permiten la oportunidad de establecer vínculos de largo plazo con estos, condición que finalmente resulta ser uno de sus factores claves para el éxito empresarial. En la práctica, los pagos basados en acciones pueden ser realizados por una empresa entregando de forma directa a sus colaboradores y suplidores de bienes y servicios acciones u opciones para la adquisición futura de acciones, lo anterior a cambio del logro de los objetivos propuestos por la empresa. Como es de suponer los objetivos cuyo cumplimiento da origen a los pagos basados en acciones son tan diversos como los son la naturaleza de cada entidad y por ende las condiciones específicas pactadas en cada uno de los acuerdos suscritos entre esta y sus empleados y proveedores. Considerando lo expuesto en los párrafos precedentes, el presente artículo tiene como objetivo abordar las generalidades del concepto de pagos basados en acciones así como el reseñar cuáles son los ámbitos en que dicho instrumento financiero es más comúnmente utilizado por los altos mandos empresariales. Como complemento de lo anterior, este artículo estudia los diferentes modelos que una entidad puede utilizar para determinar el costo financiero en que ésta debe incurrir cuando promueve un programa de pagos basados en acciones a favor de sus empleados, sus proveedores o ambos. Es así como mi estudio abarca el análisis de los diferentes métodos que existen para la valoración tanto de acciones como de las opciones que una entidad

  7. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

  8. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  9. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  10. Rotating reactors : a review

    NARCIS (Netherlands)

    Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

    2013-01-01

    This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

  11. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  12. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

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

  13. Ulysse, mentor reactor

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1981-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drivemechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displayer rods through the reactor vessel

  15. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1982-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drive mechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displacer rods through the reactor vessel. (author)

  16. Utilización de Sistemas Basados en Reglas y en Casos para diseñar transmisiones por tornillo sinfín // Use of rules based systems and cases based systems for worm gear design

    Directory of Open Access Journals (Sweden)

    Jorge Laureano Moya‐Rodríguez

    2012-01-01

    Full Text Available Las técnicas de Inteligencia Artificial se aplican hoy en día a diferentes problemas de Ingeniería,especialmente los Sistemas Basados en el Conocimiento. Entre estos últimos los más comunes son losSistemas Basados en Patrones, los Sistemas Basados en Reglas, los Sistemas Basados en Casos y losSistemas Híbridos. Los Sistemas Basados en Casos parten de problemas resueltos en un dominio deaplicación y mediante un proceso de adaptación, encuentran la solución a un nuevo problema. Estossistemas pueden ser usados con éxito para el diseño de engranajes, particularmente para el diseño detransmisiones por tornillo sin fin, sin embargo ello constituye un campo de las aplicaciones de laInteligencia Artificial aún inexplorada. En el presente trabajo se hace una comparación del uso de losSistemas Basados en Regla y los Sistemas Basados en Casos para el diseño de transmisiones portornillo sin fin y se muestran los resultados de la aplicación de los sistemas basados en regla al diseñoparticular de una transmisión por tornillo sin fin.Palabras claves: tornillo sin fin, engranajes, sistemas basados en casos, sistemas basados en reglas,inteligencia artificial.____________________________________________________________________________AbstractNowadays Artificial Intelligence techniques are applied successfully to different engineering problems,especially the “Knowledge Based Systems”. Among them the most common are the “Frame basedSystems”, “Rules Based Systems”, “Case Based Systems” and "Hybrid Systems". The “Case BasedSystems” (CBS analyze solved problems in an application domain and by means of a process ofadaptation; they find the solution to a new problem. These systems can be used successfully for thedesign of gears, particularly for designing worm gears; nevertheless it constitutes a field of the applicationsof artificial intelligence even unexplored. A comparison of the use of “Rules Based System” and

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

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

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

  18. Evaluation, Comparison and Optimization of the Compact Recuperator for the High Temperature Gas-Cooled Reactor (HTGR) Helium Turbine System

    International Nuclear Information System (INIS)

    Hao Haoran; Yang Xiaoyong; Wang Jie; Ye Ping; Yu Xiaoli; Zhao Gang

    2014-01-01

    Helium turbine system is a promising method to covert the nuclear power generated by the High Temperature Gas Cooled Reactor (HTGR) into electricity with inherent safety, compact configuration and relative high efficiency. And the recuperator is one of the key components for the HTGR helium turbine system. It is used to recover the exhaust heat out of turbine and pass it to the helium from high pressure compressor, and hence increase the cycle’s efficiency dramatically. On the other hand, the pressure drop within the recuperator will reduce the cycle efficiency, especially on low pressure side of recuperator. It is necessary to optimize the design of recuperator to achieve better performance of HTGR helium turbine system. However, this optimization has to be performed with the restriction of the size of the pressure vessel which contains the power conversion unit. This paper firstly presents an analysis to investigate the effects of flow channel geometry, recuperator’s power and size on heat transfer and pressure drop. Then the relationship between the recuperator design and system performance is established with an analytical model, followed by the evaluations of the current recuperator designs of GT-MHR, GTHTR300 and PBMR, in which several effective technical measures to optimize the recuperator are compared. Finally it is found that the most important factors for optimizing recuperator design, i.e. the cross section dimensions and tortuosity of flow channel, which can also be extended to compact intermediate heat exchangers. It turns out that a proper optimization can increase the cycle’s efficiency by 1~2 percentage, which could also raise the economy and competitiveness of future commercial HTGR plants. (author)

  19. Measurement channel of neutron flow based on software; Canal de medicion de flujo neutronico basado en software

    Energy Technology Data Exchange (ETDEWEB)

    Rivero G, T.; Benitez R, J. S. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: trg@nuclear.inin.mx

    2008-07-01

    The measurement of the thermal power in nuclear reactors is based mainly on the measurement of the neutron flow. The presence of these in the reactor core is associated to neutrons released by the fission reaction of the uranium-235. Once moderate, these neutrons are precursors of new fissions. This process it is known like chain reaction. Thus, the power to which works a nuclear reactor, he is proportional to the number of produced fissions and as these depend on released neutrons, also the power is proportional to the number of present neutrons. The measurement of the thermal power in a reactor is realized with called instruments nuclear channels. To low power (level source), these channels measure the individual counts of detected neutrons, whereas to a medium and high power, they measure the electrical current or fluctuation of the same one that generate the fission neutrons in ionization chambers especially designed to detect neutrons. For the case of TRIGA reactors, the measurement channels of neutron flow use discreet digital electronic technology makes some decades already. Recently new technological tools have arisen that allow developing new versions of nuclear channels of simple form and compacts. The present work consists of the development of a nuclear channel for TRIGA reactors based on the use of the correlated signal of a fission chamber for ample interval. This new measurement channel uses a data acquisition card of high speed and the data processing by software that to the being installed in a computer is created a virtual instrument, with what spreads in real time, in graphic and understandable form for the operator, the power indication to which it operates the nuclear reactor. This system when being based on software, offers a major versatility to realize changes in the signal processing and power monitoring algorithms. The experimental tests of neutronic power measurement show a reliable performance through seven decades of power, with a

  20. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

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

  1. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

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

  2. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  3. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    Akatsu, Eiko

    1981-12-01

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  4. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

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

  5. The fast breeder reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

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

  7. Reactor power control device

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  8. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

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

  9. The Maple reactor project

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  10. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

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

  11. Selección de Canales en Sistemas BCI basados en Potenciales P300 mediante Inteligencia de Enjambre

    Directory of Open Access Journals (Sweden)

    V. Martínez-Cagigal

    2017-10-01

    Full Text Available Resumen: Los sistemas Brain-Computer Interface (BCI se definen como sistemas de comunicación que monitorizan la actividad cerebral y traducen determinadas características, correspondientes a las intenciones del usuario, en comandos de control de un dispositivo. La selección de canales en los sistemas BCI es fundamental para evitar el sobre-entrenamiento del clasificador, reducir la carga computacional y aumentar la comodidad del usuario. A pesar de que se han desarrollado varios algoritmos con anterioridad para tal fin, las metaheurísticas basadas en inteligencia de enjambre aún no han sido suficientemente explotadas en los sistemas BCI basados en potenciales P300. En este estudio se muestra una comparativa entre cinco métodos de enjambre, basados en el comportamiento de sistemas biológicos, aplicados con el objetivo de optimizar la selección de canales en este tipo de sistemas. Los métodos se han evaluado sobre la base de datos de la “III BCI Competition 2005”, reportando precisiones similares o, en algunos casos, incluso más altas que las obtenidas sin realizar ningún tipo de selección. Dado que los cinco métodos se han demostrado capaces de disminuir drásticamente los 64 canales originales a menos de la mitad sin comprometer el rendimiento del sistema, así como de superar el conjunto típico de 8 canales y el método backward elimination, se concluye que todos ellos son adecuados para su aplicación en la selección de canales en sistemas P300-BCI. Abstract: Brain-Computer Interfaces (BCI are direct communication pathways between the brain and the environment that translate certain features, which correspond to users’ intentions, into device control commands. Channel selection in BCI systems is essential to avoid over-fitting, to reduce the computational cost and to increase the users’ comfort. Although several algorithms have previously developed for that purpose

  12. Tratamientos basados en la evidencia para adolescentes con trastornos por consumo de cannabis en el Sistema Público de Salud

    Directory of Open Access Journals (Sweden)

    Sergio Fernández Artamendi

    2014-01-01

    Full Text Available El objetivo de este estudio era describir la implementación de dos programas basados en la evidencia (PBE para adolescentes con trastornos por consumo de cannabis en el Sistema Público de Salud, y sus principales resultados. La Aproximación de Reforzamiento Comunitario para Adolescentes (A-CRA y el Control de Contingencias (MC fueron elegidos como los programas de intervención más eficaces para esta población. Un total de 26 adolescentes participaron en el estudio (91,7% chicos; edad media = 16,5 años en dos centros de carácter ambulatorio en España. Se utilizó un diseño cuasi-experimental, donde un grupo recibió A-CRA y el otro A-CRA+MC. La implementación de ambos programas resultó factible, con resultados clínicos positivos. El A-CRA ofreció buenas tasas de retención (81,3% y abstinencia (68,6%. Los resultados del grupo A-CRA+MC no fueron significativamente mejores que los del A-CRA en retención (100% o abstinencia (75,5%, aunque el limitado tamaño muestral no permite establecer conclusiones firmes. Los problemas asociados al cannabis y la sintomatología depresiva se redujeron durante el tratamiento. Varias limitaciones nos impiden determinar la eficacia clínica del A-CRA en este estudio. El proceso de traslación de los PBE al contexto clínico presentó múltiples dificultades que deben ser abordadas. Se discuten recomendaciones para futuros intentos de implementación de PBE en estos contextos.

  13. Sistema experto basado en lógica difusa tipo 1 para determinar el grado de riesgo de preeclampsia

    Directory of Open Access Journals (Sweden)

    Edna Rocio Núñez Flórez

    2014-01-01

    Full Text Available La preeclampsia es una enfermedad que pueden desarrollar las mujeres en estado de embarazo, y según el DANE es responsable del 35 % de las muertes maternas en Colombia. Ante esta situación, el objetivo de este artículo es presentar un sistema experto (SE basado en lógica difusa tipo I que permite identificar el nivel de riesgo de sufrir la enfermedad, y posibilita un diagnóstico precoz y la vigilancia estricta de la mujer gestante, aspectos fundamentales para prevenir las complicaciones asociadas a la preeclampsia. Para llevar a cabo la investigación se realizó la revisión bibliográfica para conocer los factores de riesgo que generan la enfermedad; con apoyo de un médico se establecieron los factores que se deben tener en cuenta (variables de entrada y la base de reglas, los cuales son los componentes principales del SE. Posteriormente se realiza la implementación del software con las herramientas MySql como base de datos y Java como lenguaje de programación. Para la validación de tomaron 30 historias clínicas suministradas por la Secretaría de Salud Departamental del Caquetá. El resultado del SE arrojó un 94.17 % de efectividad con un margen de error del 5.83 %, comparados con los resultados proporcionados por el especialista que analizó las historias clínicas.

  14. Cinemática inversa de robot serial utilizando algoritmo genético basado en MCDS

    Directory of Open Access Journals (Sweden)

    Juan Jairo Vaca González

    2015-04-01

    Full Text Available Los robots manipuladores seriales son herramientas eficaces para realizar tareas repetitivas y de precisión en la industria, siempre que se comprenda la cinemática involucrada en el posicionamiento y orientación del efector final. Este artículo presenta una metodología para resolver el problema cinemático inverso de un robot serial (Melfa RV-2A utilizando un algoritmo genético (AG a partir del modelo cinemático directo Screws (MCDS. Para esto, se obtienen los parámetros Screw que modelan el robot, se calcula el espacio de trabajo asociado y se diseña el AG contemplando una función multi-objetivo de alcance de posición y orientación en que se sitúa el efector final, con respecto a una coordenada y orientación de un punto objetivo establecido. La validación del AG se realiza según la aptitud, el tiempo de convergencia y la cantidad de generaciones usadas por la función para alcanzar el objetivo. Por tanto, la implementación de un AG basado en un MCDS es una herramienta prometedora que podría utilizarse para calcular la cinemática inversa de robots seriales. Esta novedosa implementación permite establecer por primera vez la exposición matricial de un sistema cinemático directo para obtener la solución cinemática inversa de un robot serial. En consecuencia, se demuestra que esta es una metodología factible y eficiente para solucionar la cinemática inversa de cualquier tipo de robot manipulador.

  15. APRENDIZAJE BASADO EN PROBLEMAS PARA DESARROLLAR ALFABETIZACIÓN CRÍTICA Y COMPETENCIAS CIUDADANAS EN EL NIVEL ELEMENTAL

    Directory of Open Access Journals (Sweden)

    Aura González Robles

    2016-01-01

    Full Text Available Este artículo presenta los hallazgos de una investigación-acción que se realizó en el tercer grado de una escuela pública en Puerto Rico en un curso que integra las disciplinas de Artes del Lenguaje, Ciencias y Estudios Sociales. Este estudio cualitativo exploró en qué medida el Aprendizaje Basado en Problemas (ABP facilita el desarrollo de la alfabetización crítica y las competencias ciudadanas. Las técnicas de recopilación de información consistieron en observaciones participativas y no-participativas, notas de campo, fotografías y materiales que desarrollaron los estudiantes. Teóricamente, este trabajo se fundamenta en el constructivismo social, en particular las ideas de Paulo Freire, así como en los conceptos de competencias ciudadanas y alfabetización ciudadana como fundamentos para el desarrollo humano integral. El análisis de los datos permite concluir que el ABP fomenta, a través de sus distintos pasos, el desarrollo inicial de los atributos de la alfabetización crítica y las competencias ciudadanas de las niñas y los niños de tercer grado. De acuerdo con los hallazgos se observó que el atributo de recursos personales y culturales, correspondiente a la competencia de alfabetización crítica se propició en todos los pasos de ABP. Asimismo, el atributo más frecuente relacionado a las competencias ciudadanas fue la capacidad de inquirir.

  16. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  17. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

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

  18. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

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

  19. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  20. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

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

  1. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

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

  2. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

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

  3. Efecto de un programa de entrenamiento físico basado en la secuencia de desarrollo sobre el balance postural en futbolistas: ensayo controlado aleatorizado

    Directory of Open Access Journals (Sweden)

    Erika Mancera Soto

    2013-10-01

    Full Text Available Antecedentes. En la actividad deportiva, el balance postural es requerido para mantener la estabilidad durante el juego. Por tanto, existe una necesidad de determinar si el aprendizaje motor desde posiciones funcionales, movimientos coordinados están implicados en los deportes. Objetivo. El objetivo de este estudio fue determinar el efecto de un entrenamiento físico basado en la secuencia de desarrollo sobre el balance postural en futbolistas de la selección de la Universidad Nacional de Colombia, sede Bogotá. Materiales y métodos. Ensayo controlado aleatorizado de 19 hombres adultos jóvenes pertenecientes al equipo de futbol de la Universidad Nacional de Colombia, sede Bogotá. Los participantes fueron aleatorizados y asignados a dos grupos, un grupo intervención (n=11, en el cual se le aplicó un entrenamiento físico basado en la secuencia de desarrollo y un grupo control (n=8 el cual realizó un programa de entrenamiento convencional de futbol. Resultados. Existe homogeneidad entre los dos grupos, en el test de balance dinámico SEBT los valores obtenidos (P<0,5, demuestran una mejoría en todas las direcciones evaluadas tanto en el miembro inferior derecho como en el izquierdo. En el grupo de intervención, la relación intragrupal muestra una correlación 3:1 siendo una medida de protección. Conclusión. La aplicación de un programa de entrenamiento físico basado en la secuencia de desarrollo genera importantes mejoras en el balance estático y dinámico. Se demuestran mejoras en las distancias de excursión lo cual se puede relacionar con un aumento del control postural dinámico.

  4. Desarrollo de la competencia de aprendizaje autónomo en estudiantes de Pedagogía en un modelo educativo basado en competencias

    Directory of Open Access Journals (Sweden)

    Malva Lidia Reyes Roa

    2017-01-01

    Full Text Available En el ámbito de la Educación Superior se han producido cambios y renovaciones entre estos destaca la implementación de modelos basados en competencias en la formación pro - fesional. No obstante, a pesar de haber sido implementado, son pocos los estudios que evalúen las variaciones de las competencias básicas como el aprendizaje autónomo. El artículo describe el desarrollo de la competencia de aprendizaje autónomo que alcanzan estudiantes universita - rios de pedagogía (educación: parvularia, básica, básica intercultural y diferencial de primer y tercer año correspondientes a las cohortes 2012 y 2010 esta última la primera generación en un Modelo Educativo basado en Competencias. Participaron 254 estudiantes. Se utilizó un diseño descriptivo comparativo. Los resultados indican que las diferencias existentes se dan en aspectos específicos de actuación evidenciados en indicadores entre los que destacan aquellos orientados a la gestión del tiempo vinculados a mostrar cómo el estudiante adecua su propia calendarización a las necesidades de aprendizaje propuestas por el docente ( p = 0,002 y planifica para la resolu - ción de problemas emergentes ( p = 0,001. Sin embargo, no existen diferencias significativas en las dimensiones que componen este constructo ( p =0.096. Aun cuando en todos los casos estas diferencias son favorables a los estudiantes de tercer año ( M =3,57 sobre primer año ( M =3,27. El artículo concluye proyectando los desafíos de un modelo basado en competencia sobre el de - sarrollo de componentes básicos que complementan el aprendizaje autónomo.

  5. Modelos basados en agentes: aportes epistemológicos y teóricos para la investigación social

    Directory of Open Access Journals (Sweden)

    Leonardo Gabriel Rodríguez Zoya

    2015-01-01

    Full Text Available Los modelos basados en agentes (MBA constituyen una nueva generación de métodos computacionales que permiten modelar la estructura de un sistema complejo y simular su evolución dinámica a lo largo deI tiempo. EI uso de los MBA constituye una tendencia metodológica en expansión en las ciencias sociales contemporáneas; sin embargo, continúan siendo poco conocidos y enseiíados en el campo sociológico, de modo que constituyen una alternativa metodológica minoritaria entre los investigadores sociales. EI propósito de este trabajo es introducir a los científicos sociales en las ideas centrales de los modelos basados en agentes a partir de su articulación con ciertos problemas teóricos y metodológicos cruciales de las ciencias sociales. La primera sección problematiza la relación entre los modelos basados en agentes y los sistemas complejos en una perspectiva epistemológica crítica. Posteriormente, se analizan los aportes de los MBA a la investigación social, y en la tercera sección se evalúa críticamente su aplicación en el marco de una disciplina particular: la ciencia política. Finalmente, se desarrolla un ejemplo práctico de una simulación basada en agentes a partir del trabajo clásico de Thomas Schelling sobre segregación racial.

  6. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

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

  7. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  8. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

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

  9. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1980-09-01

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

  11. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

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

  12. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

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

  13. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

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

  14. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  15. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  16. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  17. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  18. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  19. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  20. The Dragon reactor experiment

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

  1. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  2. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  3. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics

  4. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  5. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  6. Preparación y caracterización de filmes epoxi basados en aceite de linaza para aplicación

    OpenAIRE

    Gómez Carrera, Marta

    2015-01-01

    En este proyecto se preparan diferentes composiciones de filmes epoxi basados en aceite de linaza químicamente epoxidado (Epoxidized Linseed Oil-ELO) y su posterior curado con dos tipos de endurecedores, una poliaminoamida comercial y una poli(etilenimina) con dos pesos moleculares distintos. El objetivo final es la preparación de una pintura epoxi menos tóxica para el ser humano, sustituyendo parcialmente la molécula de diglicidil éter de bisfenol A (DGEBA), de las formulaciones epoxi, por m...

  7. Propuesta de utilización del razonamiento basado en casos para la recuperación de procedimientos de prueba funcionales

    Directory of Open Access Journals (Sweden)

    Martha Dunia Delgado Dapena

    2010-09-01

    Full Text Available En este trabajo se presenta una propuesta de estructura de almacenamiento y los mecanismos de recuperación utilizados para aplicar el razonamiento basado en casos (RBC en la generación de procedimientos de prueba funcionales en proyectos de software. Esta propuesta parte de los requisitos funcionales del proyecto de software y en ella se enuncian los algoritmos propuestos para considerar la semejanza entre cada par de proyectos, así como los que permiten adaptar la solución encontrada en la base de casos a las características de los nuevos proyectos.

  8. Desarrollo competencial en educación infantil a través del aprendizaje basado en proyectos en centros educativos de jaén

    OpenAIRE

    Palomares Ruiz, Pedro

    2017-01-01

    Esta tesis surge derivada de la incertidumbre generada acerca de la idoneidad de la implementación de la práctica pedagógica, aprendizaje basado en proyectos. Tras constatar que a pesar de ser una práctica muy conocida no era proporcional a su grado de implementación suscita el interés por conocer el grado competencial adquirido por el alumnado en el tercer nivel de educación infantil, de los centros de la provincia de Jaén, en los que se ha implementado la práctica pedagógica objeto de estud...

  9. Desarrollo de un controlador para motores DC brushless basado en CompactRIO y LabVIEW de National Instruments para el estudio de nuevos algoritmos de control

    OpenAIRE

    García Haro, Juan Miguel

    2011-01-01

    Este proyecto que el CAR inició hace pocos años tiene como objetivo principal el estudio y desarrollo de nuevas tecnologías en el campo de actuación y control automático, que servirá de base para otras futuras investigaciones dentro del centro. La tecnología a la que se hace mención se refiere al control de actuadores basados en motores DC brushless (BLDC Motors) empleando el sistema de hardware embebido CompactRIO y programación LabVIEW de National Instruments. Tradicionalmente se emplea en ...

  10. RETOS Y DESAFÍOS DE LOS DOCENTES DE EDUCACIÓN SECUNDARIA EN LA IMPLEMENTACIÓN DEL ENFOQUE BASADO EN COMPETENCIAS

    OpenAIRE

    Toalá Valdez, Adriana del Carmen

    2015-01-01

    La presente investigación tuvo como principal objetivo analizar los retos y desafíos para los docentes de educación secundaria en la implementación del enfoque basado en competencias. La introducción de este enfoque en la educación básica mexicana y española, así como en otros niveles educativos ha sido causa de controversias y desacuerdos, dada las opiniones de investigadores y especialistas que consideran que únicamente se pretende formar gente que satisfaga las necesidades del mercado labo...

  11. UNA APROXIMACIÓN A LA PLANEACIÓN MINERA A CIELO ABIERTO DESDE UN ENFOQUE BASADO EN DECISIONES BAJO INCERTIDUMBRE

    OpenAIRE

    FRANCO SEPÚLVEDA, GIOVANNI; BRANCH BEDOYA, JOHN WILLIAN; JARAMILLO ÁLVAREZ, PATRICIA

    2010-01-01

    En este artículo se presenta en primer lugar una descripción del proceso de planeamiento minero desde un punto de vista clásico y su relación con los actuales procesos de planeamiento minero que tienen como base la optimización en sus diferentes etapas. Seguidamente, se realiza un acercamiento a los procesos de planeación minera a cielo abierto desde un enfoque basado en decisiones bajo incertidumbre. Por último, se lleva a cabo un análisis crítico de los artículos encontrados clasificándolos...

  12. Una aproximación a la planeación minera a cielo abierto desde un enfoque basado en decisiones bajo incertidumbre

    OpenAIRE

    FRANCO SEPÚLVEDA, GIOVANNI; BRANCH BEDOYA, JOHN WILLIAN; JARAMILLO ÁLVAREZ, PATRICIA

    2011-01-01

    En este artículo se presenta en primer lugar una descripción del proceso de planeamiento minero desde un punto de vista clásico y su relación con los actuales procesos de planeamiento minero que tienen como base la optimización en sus diferentes etapas. Seguidamente, se realiza un acercamiento a los procesos de planeación minera a cielo abierto desde un enfoque basado en decisiones bajo incertidumbre. Por último, se lleva a cabo un análisis crítico de los artículos encontrados clasificándolos...

  13. Un algoritmo de clasificación incremental basado en los k vecinos más similares para datos mezclados

    Directory of Open Access Journals (Sweden)

    Guillermo Sánchez-Díaz

    2013-01-01

    Full Text Available En este trabajo, se presenta un algoritmo de clasificación incremental basado en los k vecinos más similares, el cual permite trabajar con datos mezclados y funciones de semejanza que no necesariamente son distancias. El algoritmo presentado es adecuado para procesar grandes conjuntos de datos, debido a que sólo almacena en la memoria principal de la computadora los k vecinos más similares procesados hasta el paso t, recorriendo una sola vez el conjunto de datos de entrenamiento. Se presentan resultados obtenidos con diversos conjuntos de datos sintéticos y reales.

  14. Sistema de gestión en seguridad basado en la norma OHSAS 18001 para la empresa constructora eléctrica IELCO

    OpenAIRE

    Bustamante Granda, Fernando

    2013-01-01

    La presente tesis es el resultado de un minucioso estudio, desarrollado con el principal objetivo de proponer un sistema de gestión de Seguridad y Salud Ocupacional, basado en la OHSAS 18001:2007, para la empresa Constructora Eléctrica IELCO y así ayudar a mejorar la seguridad y la salud ocupacional de los trabajadores de la empresa. La investigación se realizó tanto en el campo, lugar donde se desarrollaban los proyectos de construcciones de Redes de Distribución Eléctrica, como en las o...

  15. NUEVO MODELO DIDACTICO CURRICULAR PARA LA ENSEÑANZA DE LA CIENCIA, BASADO EN LA ACUICULTURA DESTINADA A ENSEÑANZA BASICA Y MEDIA, EN ESTABLECIMIENTOS COSTEROS

    OpenAIRE

    Toledo Muñoz, Héctor Eladio

    2013-01-01

    El Proyecto “Nuevo modelo didáctico curricular para la enseñanza de la ciencia, basado en la acuicultura destinada a la Enseñanza Básica y Media en establecimientos costeros”. Se desarrolló durante el periodo 2007 – 2010, con la participación de colegios contrapartes y experimentales de la Región de Los Lagos, con un espectro socio-educacional del área particular, particular subvencionada y municipalizada. La metodología utilizada consistió básicamente en el desarrollo de talleres, en el c...

  16. Guía de implantación de un SGSI basado en la norma UNE-ISO/IEC 27001

    OpenAIRE

    Muñoz Martín, Manuel

    2015-01-01

    Guía para el análisis, desarrollo e implantación de un sistema de gestión de la seguridad de la información basado en la norma ISO 27001 en una empresa tecnológica. Guia per a l'anàlisi, desenvolupament i implantació d'un sistema de gestió de la seguretat de la informació basat en la norma ISO 27001 en una empresa tecnològica. Bachelor thesis for the Computer Science program.

  17. Combinación entre Algoritmos Genéticos y Aleatorios para la Programación de Horarios de Clases basado en Ritmos Cognitivos

    OpenAIRE

    Castrillón, Omar D

    2014-01-01

    Se ha diseñado un método basado en algoritmos evolutivos (genéticos y aleatorios) para programar los horarios de clases en una universidad. Esta metodología considera los ritmos cognitivos de los estudiantes que indican que es mejor enseñar algunas asignaturas en intervalos específicos de tiempo. Primero se describen las diferentes técnicas empleadas para desarrollar este problema. Luego se propone una nueva metodología basada en ritmos cognitivos y algoritmos evolutivos, para resolver todas ...

  18. Estrategias para trabajar la creatividad en la Educación Superior: pensamiento de diseño, aprendizaje basado en juegos y en proyectos.

    OpenAIRE

    González González, Carina Soledad

    2015-01-01

    Existen prácticas docentes que estimulan una mayor participación de los estudiantes, dando lugar a un trabajo motivador que estimula el pensamiento creativo e innovador, que potencia su autonomía y facilita el aprendizaje de competencias transversales y profesionales. En este trabajo, presentamos una experiencia de innovación educativa en la enseñanza de la ingeniería y del diseño, en donde se han aplicado estrategias de enseñanza- aprendizaje basada en proyectos (PBL) y aprendizaje basado en...

  19. Sistema de diálogo basado en mensajería instantánea para el control de dispositivos en el Internet de las cosas

    OpenAIRE

    Noguera Arnaldos, Jose Angel

    2016-01-01

    1. OBJETIVOS DE LA TESIS Desarrollar y sentar las bases de un sistema de comunicación basado en redes sociales en formato chat para que los humanos nos podamos comunicar en lenguaje natural y en tiempo real con las máquinas. 2. METODOLOGÍA Para lograr este objetivo, se ha seguido la siguiente metodología: • Análisis del estado del arte en Internet de las cosas, sistemas de diálogo y agentes conversacionales, inteligencia ambiental y domótica, bases de conocimiento y ontologías...

  20. Sistema predictivo basado en aprendizaje automático para la deserción estudiantil en instituciones de educación superior

    OpenAIRE

    Pájaro Fuentes, Leandro

    2016-01-01

    El objetivo principal del trabajo es desarrollar un sistema predictivo basado en aprendizaje automático para evitar la deserción estudiantil en instituciones de educación superior. L'objectiu principal del treball és desenvolupar un sistema predictiu basat en aprenentatge automàtic per evitar la deserció escolar en institucions d'educació superior. The main objective of the work is to develop a predictive system based on automatic learning to avoid student dropout in institutions of hig...