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Sample records for thales thermohydraulic loca

  1. Flexure bearing cryocoolers at Thales Cryogenics

    Science.gov (United States)

    Meijers, M.; Benschop, A. A. J.; Mullié, J. C.

    2002-05-01

    Thales Cryogenics (NL) and Thales Cryogenie (F), formerly known as Signal Usfa and Cryotechnologies, closely co-operate in the field of production and development of linear and rotary cryocoolers. Over the past years, Thales Cryogenics has developed a complete range of Stirling cryocoolers with flexure bearings. In this paper the main design features of the flexure bearing compressor are explained. With these flexure bearing cryocoolers, which are available in slip-on configuration as well as IDCA (Integrated Detector Cooler Assembly), up to 6 W @80 K cooling power can be obtained. Also a pulse tube cryocooler with a specified cooling power of 500 mW @80 K has been developed. Two specific production machines have been developed and introduced in the production line. With this equipment Thales Cryogenics has been able to further improve the quality and reproducibility of its coolers. Up to now, several flexure bearing cryocoolers have been built and integrated in various new commercial and military applications requiring long life cryocoolers. Besides this, Thales Cryogenics is active in several space applications in co-operation with Air Liquide/DTA.

  2. Multiplication: From Thales to Lie1

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 17; Issue 5. Multiplication: From Thales to Lie. Pradipkumar H Keskar. General Article Volume 17 Issue 5 May 2012 ... Author Affiliations. Pradipkumar H Keskar1. Mathematics Department, Birla Institute of Technology and Science, Pilani 333 031, India.

  3. The nature of water: Thales' arkhe.

    Science.gov (United States)

    De Santo, Natale G; Bisaccia, Carmela; Bilancio, Giancarlo; Romano, Mercedes; Cirillo, Massimo

    2009-01-01

    Thales was born into a noble family of Phoenician origin at the time of the 25th Olympiad (floruit 585 bc; he was 40 in the year of the solar eclipse. He had no teachers but had occasion to learn from Egyptian priests. He developed into a scholar and politician very much appreciated by Heraclitus, Herodotus and Democritus, and was always considered a man of practical wisdom. He was probably the first to speak about the immortality of the soul. He is listed as the first of many unmarried men who paved the road for philosophy. For Diogenes Laertius (Lives and Opinions of Eminent Philosophers), he was the instructor of Anaximander. Thales, the man who first discovered how to draw a right-angle triangle in a circle, was the first philosopher of nature (physis). "Philosophy begins with Thales," pointed out Bertrand Russell in 1961. This honor had been conceded also by Aristotle: "Anaximander, Thales' pupil, founded the Ionian tradition of philosophy." Many explanations may be given for the importance of water, including its importance for living processes, the economic role of the Nile, the importance of the port for Miletus and the fact that Ocean and Thetys were in Homer's tradition progenitors of the world.

  4. Thales and the Dawn of Western Philosophy

    Science.gov (United States)

    Morano, Donald V.

    1975-01-01

    From the insights of ancient and modern commentators and historical and psychological research, the author strived to determine briefly: (1) what Thales meant by his statements, (2) how he came to such conclusions, and (3) why such conclusions are now regarded as marking the dawn of Western philosophy. (Author/RK)

  5. The Origins of Science Thales' Leap

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 1; Issue 4. The Origins of Science Thales' Leap. Gangan Prathap. Reflections Volume 1 Issue 4 April 1996 pp 67-73. Fulltext. Click here to view fulltext PDF. Permanent link: http://www.ias.ac.in/article/fulltext/reso/001/04/0067-0073. Author Affiliations.

  6. Assessment of the Thermal Hydraulic Models in THALES

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Byeong Il; Kim, Hong Ju; Jang, Beomjun; Woo, Hae-Seuk [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2016-10-15

    THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) developed by KEPCO Nuclear Fuel is a subchannel analysis code on the basis of the single-stage core analysis model. THALES calculates the local fluid conditions and DNBR in the PWR (Pressurized Water Reactor) core. Currently, THALES is limited to the licensed type of the nuclear power plant because the thermal hydraulic models and CHF (Critical Heat Flux) correlations for OPR1000 and APR1400 are only licensed. KEPCO NF intends to apply THALES to WH typed nuclear power plants in Korea. To expand the applicable types of the nuclear power plants, the existing thermal hydraulic models were modified and new thermal hydraulic models were added to THALES. In this study, the thermal hydraulic models tested and added in THALES are reviewed and a preliminary calculation is performed. KEPCO NF intends to apply THALES to various typed nuclear power plants in Korea. So, the existing thermal hydraulic models implemented in THALES are modified and the void model which are generally used in the subchannel analysis code is added. Through the preliminary calculation, it is confirmed that the thermal hydraulic models are properly modified and implemented in THALES, which shows the possibility to apply THALES in various typed nuclear power plants in Korea.

  7. Implementation of Boundary Condition to THALES Code

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Beomjun; Chun, Chong Kuk; Park, Ho Young; Woo, Hae-Seuk [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2016-10-15

    The boundary condition of momentum equation of THALES code utilizes the exit pressure boundary to solve the elliptic partial difference momentum equations. This method is the same as the most of the subchannel analysis codes. Other codes such as VIPRE utilize the uniform pressure distribution as outlet boundary condition. In this case, uniform inlet flow rate is assumed. In order to test the core flow field regarding the boundary conditions, analysis was performed for two core conditions. One condition is nominal plant operating condition. In this paper, generic THALES power distribution is used. For nominal operation case, there are no different results depending on the type of outlet pressure boundary condition. But low-power and high-peaking case, density difference for lateral direction becomes large due to high peaking power of core. Since density change causes pressure change, In this case, uniform outlet pressure distribution can't be assumed anymore. Design outlet pressure distribution is measured at nominal core condition. Therefore, design outlet pressure distribution also can't be used due to the difference in core power and flow rate. As a result, it is reasonable that neumann boundary condition is applied in low-power and high peaking core condition including various accident condition.

  8. Thales van Milete en het begin van de filosofie

    NARCIS (Netherlands)

    Poorthuis, Marcel

    2016-01-01

    Het water als oorsprong van alles is niet alleen een natuurkundige of biologische gedachte, maar tevens een metafysisische idee over de oorsprong van alle dingen. Deze uitspraak van Thales van Milete markeert het begin van de filosofie. Zowel kerkvaders als moderne filosofen onder wie Nietzsche en

  9. WEAG THALES JP11.20 (REVVA) Results and Perspectives

    NARCIS (Netherlands)

    Jaquart, R.; Brade, D.; Voogd, J.M.; Yi, C.H.

    2005-01-01

    Within the WEAG THALES Joint Program 11.20 "Common Framework for Verification, Validation, and Accreditation of Simulations" (nicknamed "REVVA") between Denmark, France, Italy, Sweden, and The Nether-lands, a new customer-based and product-oriented VV&A methodology was developed. It includes: (1)

  10. Estimation of LOCA break size using cascaded Fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA), which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict the progression of LOCAs in nuclear power plants. Therefore, accurate information on the LOCA break position and size should be provided to efficiently manage the accident. In this paper, the LOCA break size is predicted using a cascaded fuzzy neural network (CFNN) model. The input data of the CFNN model are the time-integrated values of each measurement signal for an initial short-time interval after a reactor scram. The training of the CFNN model is accomplished by a hybrid method combined with a genetic algorithm and a least squares method. As a result, LOCA break size is estimated exactly by the proposed CFNN model.

  11. Estimation of LOCA Break Size Using Cascaded Fuzzy Neural Networks

    Directory of Open Access Journals (Sweden)

    Geon Pil Choi

    2017-04-01

    Full Text Available Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA, which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict the progression of LOCAs in nuclear power plants. Therefore, accurate information on the LOCA break position and size should be provided to efficiently manage the accident. In this paper, the LOCA break size is predicted using a cascaded fuzzy neural network (CFNN model. The input data of the CFNN model are the time-integrated values of each measurement signal for an initial short-time interval after a reactor scram. The training of the CFNN model is accomplished by a hybrid method combined with a genetic algorithm and a least squares method. As a result, LOCA break size is estimated exactly by the proposed CFNN model.

  12. Drawing physics 2600 year of discovery from Thales to Higgs

    CERN Document Server

    Lemons, Don S

    2017-01-01

    Humans have been trying to understand the physical universe since antiquity. Aristotle had one vision (the realm of the celestial spheres is perfect), and Einstein another (all motion is relativistic). More often than not, these different understandings begin with a simple drawing, a pre-mathematical picture of reality. Such drawings are a humble but effective tool of the physicist's craft, part of the tradition of thinking, teaching, and learning passed down through the centuries. This book uses drawings to help explain fifty-one key ideas of physics accessibly and engagingly. Don Lemons, a professor of physics and author of several physics books, pairs short, elegantly written essays with simple drawings that together convey important concepts from the history of physical science. Lemons proceeds chronologically, beginning with Thales' discovery of triangulation, the Pythagorean monocord, and Archimedes' explanation of balance. He continues through Leonardo's description of -earthshine- (the ghostly glow b...

  13. Miniature cryocooler developments for high operating temperatures at Thales Cryogenics

    Science.gov (United States)

    Arts, R.; Martin, J.-Y.; Willems, D.; Seguineau, C.; Van Acker, S.; Mullié, J. C.; Göbel, A.; Tops, M.; Le Bordays, J.; Etchanchu, T.; Benschop, A. A. J.

    2015-05-01

    In recent years there has been a drive towards miniaturized cooled IDCA solutions for low-power, low-mass, low-size products (SWaP). To support this drive, coolers are developed optimized for high-temperature, low heat load dewar-detector assemblies. In this paper, Thales Cryogenics development activities supporting SWaP are presented. Design choices are discussed and compared to various key requirements. Trade-off analysis results are presented on drive voltage, cold finger definition (length, material, diameter and sealing concept), and other interface considerations, including cold finger definition. In parallel with linear and rotary cooler options, designs for small-size high-efficiency drive electronics based on state-of-the-art architectures are presented.

  14. Locally Constructed Analogs (LOCA) Daily CMIP5 Climate Projections

    Data.gov (United States)

    US Bureau of Reclamation, Department of the Interior — This archive contains 64 projections of daily LOCA CMIP5 climate over the contiguous United States. This archive includes these meteorological variables at 1/16th...

  15. Thermal performance testing of two Thales 9310 pulse-tube cryocoolers for PHyTIR

    Science.gov (United States)

    Paine, Christopher G.

    2014-01-01

    PHyTIR is a NASA-funded technology demonstration for a near-term earth-observing instrument in the thermal infrared spectrum, intended for use in the HyspIRI mission. PHyTIR will use two Thales 9310 single-stage pulse tube cryocoolers, one to directly cool the FPA, the other to simulate a passive radiator. We report performance measurements for the two Thales 9310 cryocoolers intended for inclusion in the PHyTIR demonstrator.

  16. An IPSN research programme to resolve pending LOCA issues

    Energy Technology Data Exchange (ETDEWEB)

    Mailliat, A.; Grandjean, C.; Clement, B. [CEA Cadarache, Inst. de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    Studies performed in IPSN and elsewhere pointed out that high burnup may induce specific effects under LOCA conditions, especially those related with fuel relocation. Uncertainties exist regarding how much these effects might affect the late evolution of the accident transient and the associated safety issues. IPSN estimates that a better knowledge of specific phenomena is required in order to resolve the pending uncertainties related to LOCA criteria. IPSN is preparing the so called APRP-Irradie (High Burnup fuel LOCA) programme. One of the important aspect of this programme is in-pile experiments involving bundle geometries in the PHEBUS facility located at Cadarache, France. A feasibility study for such an experimental programme is underway and should provide soon, a finalized project including cost and schedule aspects. (authors)

  17. Radiation signature folowing the hypothesized LOCA. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.

    1977-09-01

    The study establishes the radiation source profile following the hypothesized Loss of Coolant Accident (LOCA) as suggested by the applicable Regulatory Guides. The source is specified as time-dependent gamma and beta energy release rates and energy spectra with dose and dose rate values presented for a generic containment structure. The results of the study will provide a basis for a comparison of radiation simulators used in (radiation) qualification testing of Class I components and an evaluation of simulator ''adequacy'' in duplicating the LOCA radiation environments and resultant component damage.

  18. Miniature Stirling cryocoolers at Thales Cryogenics: qualification results and integration solutions

    Science.gov (United States)

    Arts, R.; Martin, J.-Y.; Willems, D.; Seguineau, C.; de Jonge, G.; Van Acker, S.; Mullié, J.; Le Bordays, J.; Benschop, T.

    2016-05-01

    During the 2015 SPIE-DSS conference, Thales Cryogenics presented new miniature cryocoolers for high operating temperatures. In this paper, an update is given regarding the qualification programme performed on these new products. Integration aspects are discussed, including an in-depth examination of the influence of the dewar cold finger on sizing and performance of the cryocooler. The UP8197 will be placed in the reference frame of the Thales product range of high-reliability linear cryocoolers, while the rotary solution will be considered as the most compact solution in the Thales portfolio. Compatibility of the cryocoolers design with new and existing 1/4" dewar designs is examined, and potential future developments are presented.

  19. El teorema de Pitágoras y el teorema de Thales_Instrumento de evaluación desde de las Pruebas Saber / The Pythagorean Theorem and the theorem of Thales: Assessment tool from the Pruebas Saber.

    OpenAIRE

    Rangel Luengas, Juan Samuel

    2011-01-01

    Propuesta de evaluacion utilizando items tipo prueba Saber referentes a los Teoremas de Thales y Pitágoras para ello se estudian las evaluaciones externas nacionales e internacionales. / Abstract. Proposal evaluation using Prueba Saber type items concerning Theorems Thales and Pythagoras for it examines the external evaluations of national and international and proposes an evaluation tool.

  20. Minutes of the LOCA workshop meeting (Nov. 2000)

    Energy Technology Data Exchange (ETDEWEB)

    Barrera, Facundo

    2000-03-15

    A meeting was held in Halden, before the One-Hundred and Twenty-Second meeting of the Halden Programme Group in November 2000, to address the issues of a proposed series of tests related to LOCA. Several issues were defined, including the fuel segments to be used and the general test procedures. (Author)

  1. WEAG THALES JP11.20 - Final State of the REVVA Methodology

    NARCIS (Netherlands)

    Brade, D.; Jacquart, R.; Voogd, J.M.; Yi, C.H.

    2005-01-01

    Under the umbrella of the Western European Armament Group's THALES Memorandum of Under-standing, a "Common Framework for Verification, Validation, and Accreditation of Simulations" (nickname "REVVA") was developed and stabilized during the period March 2003 through October 2004. Within the Joint

  2. Thermo-hydraulic analysis of the windowless target system

    Energy Technology Data Exchange (ETDEWEB)

    Bianchi, Fosco [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy)], E-mail: fosco.bianchi@bologna.enea.it; Ferri, Roberta [SIET, Via Nino Bixio 27, 29100 Piacenza (Italy); Moreau, Vincent [CRS4, Polaris Edificio 1 CP25, 09010 Pula (Canada) (Italy)

    2008-08-15

    The target system, whose function is to supply an external neutron source to a subcritical core in order to sustain the neutron chain reaction, is the most critical part of an ADS being subject to severe thermo-mechanical loading and material damage due to accelerator protons and fission neutrons. In order to reduce the material damage and to increase the life of the target system a windowless option was chosen in the framework of the European PDS-XADS project as reference configuration for the experimental ADS cooled by lead-bismuth eutectic alloy. This document deals with the results of the thermo-hydraulic analysis performed with STAR-CD and RELAP5 codes to assess the behaviour of the windowless target system during off-normal operating conditions. It also reports a description of modifications properly implemented in the codes for studying this kind of plant. The windowless target system shows a satisfactory thermo-hydraulic behaviour for the analysed accidents, except for the loss of both pumps without proton beam shut-off and for the beam trips lasting more than 1 s.

  3. An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

    Directory of Open Access Journals (Sweden)

    Cesare Frepoli

    2008-01-01

    Full Text Available Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

  4. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  5. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [EDF/SEPTEN Villeurbanne (France); GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  6. ThaleMine: A Warehouse for Arabidopsis Data Integration and Discovery.

    Science.gov (United States)

    Krishnakumar, Vivek; Contrino, Sergio; Cheng, Chia-Yi; Belyaeva, Irina; Ferlanti, Erik S; Miller, Jason R; Vaughn, Matthew W; Micklem, Gos; Town, Christopher D; Chan, Agnes P

    2017-01-01

    ThaleMine (https://apps.araport.org/thalemine/) is a comprehensive data warehouse that integrates a wide array of genomic information of the model plant Arabidopsis thaliana. The data collection currently includes the latest structural and functional annotation from the Araport11 update, the Col-0 genome sequence, RNA-seq and array expression, co-expression, protein interactions, homologs, pathways, publications, alleles, germplasm and phenotypes. The data are collected from a wide variety of public resources. Users can browse gene-specific data through Gene Report pages, identify and create gene lists based on experiments or indexed keywords, and run GO enrichment analysis to investigate the biological significance of selected gene sets. Developed by the Arabidopsis Information Portal project (Araport, https://www.araport.org/), ThaleMine uses the InterMine software framework, which builds well-structured data, and provides powerful data query and analysis functionality. The warehoused data can be accessed by users via graphical interfaces, as well as programmatically via web-services. Here we describe recent developments in ThaleMine including new features and extensions, and discuss future improvements. InterMine has been broadly adopted by the model organism research community including nematode, rat, mouse, zebrafish, budding yeast, the modENCODE project, as well as being used for human data. ThaleMine is the first InterMine developed for a plant model. As additional new plant InterMines are developed by the legume and other plant research communities, the potential of cross-organism integrative data analysis will be further enabled. © The Author 2016. Published by Oxford University Press on behalf of Japanese Society of Plant Physiologists. All rights reserved. For permissions, please email: journals.permissions@oup.com.

  7. Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

    Directory of Open Access Journals (Sweden)

    F. Reventós

    2012-01-01

    Full Text Available Integral test facilities (ITFs are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results.

  8. Thermo-hydraulic design of earth-air heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    De Paepe, M. [Department of Flow, Heat and Combustion Mechanics, Ghent University, Ghent (Belgium); Janssens, A. [Department of Architecture and Urbanism, Ghent University, Ghent (Belgium)

    2003-07-01

    Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

  9. Thermo-hydraulic design of earth-air heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Paepe, M. de [Ghent University (Belgium). Department of Flow, Heat and Combustion Mechanics; Janssens, A. [Ghent University (Belgium). Department of Architecture and Urbanism

    2003-05-01

    Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

  10. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  11. Integrated testing of the Thales LPT9510 pulse tube cooler and the iris LCCE electronics

    Science.gov (United States)

    Johnson, Dean L.; Rodriguez, Jose I.; Carroll, Brian A.; Bustamante, John G.; Kirkconnell, Carl S.; Luong, Thomas T.; Murphy, J. B.; Haley, Michael F.

    2014-01-01

    The Jet Propulsion Laboratory (JPL) has identified the Thales LPT9510 pulse tube cryocooler as a candidate low cost cryocooler to provide active cooling on future cost-capped scientific missions. The commercially available cooler can provide refrigeration in excess of 2 W at 100K for 60W of power. JPL purchased the LPT9510 cooler for thermal and dynamic performance characterization, and has initiated the flight qualification of the existing cooler design to satisfy near-term JPL needs for this cooler. The LPT9510 has been thermally tested over the heat reject temperature range of 0C to +40C during characterization testing. The cooler was placed on a force dynamometer to measure the selfgenerated vibration of the cooler. Iris Technology has provided JPL with a brass board version of the Low Cost Cryocooler Electronics (LCCE) to drive the Thales cooler during characterization testing. The LCCE provides precision closed-loop temperature control and embodies extensive protection circuitry for handling and operational robustness; other features such as exported vibration mitigation and low frequency input current filtering are envisioned as options that future flight versions may or may not include based upon the mission requirements. JPL has also chosen to partner with Iris Technology for the development of electronics suitable for future flight applications. Iris Technology is building a set of radiation-hard, flight-design electronics to deliver to the Air Force Research Laboratory (AFRL). Test results of the thermal, dynamic and EMC testing of the integrated Thales LPT9510 cooler and Iris LCCE electronics is presented here.

  12. RMs1: qualification results of the rotary miniature Stirling cryocooler at Thales Cryogenics

    Science.gov (United States)

    Martin, Jean-Yves; Seguineau, Cédric; Van-Acker, Sébastien; Sacau, Mikel; Le Bordays, Julien; Etchanchu, Thierry; Vasse, Christophe; Abadie, Christian; Laplagne, Gilles; Benschop, Tonny

    2017-05-01

    The trend for miniaturized Integrated Dewar and Cooler Assemblies (IDCA) has been confirmed over the past few years with several mentions of a new generation of IR detector working at High Operating Temperature (HOT). This key technology enables the use of cryocooler with reduced needs of cryogenics power. As a consequence, miniaturized IDCA are the combination of a HOT IR detector coupled with a low-size, low-weight and low-power (SWaP) cryocooler. Thales Cryogenics has developed his own line of SWaP products. Qualification results on linear solution where shown last year. The current paper focuses on the latest results obtained on RMs1 prototypes, the new rotary SWaP cryocooler from Thales Cryogenics. Cryogenic performances and induced vibrations are presented. In a second part, progress is discussed on compactness and weight on one side, and on power consumption on the other side. It shows how the trade-off made between weight and power consumption could lead to an optimized solution at system level. At least, an update is made on the qualification status.

  13. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

    Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

  14. An overview of IPPE research on liquid metal fast reactor thermohydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P. [SSC RF-IPPE, Kaluga (Russian Federation)

    2003-07-01

    The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model.

  15. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  16. Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

    Directory of Open Access Journals (Sweden)

    Lázaro Lima

    2011-12-01

    Full Text Available “Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

  17. Calculation of the frequency of excedence in Full Spectrum LOCA by FSR; Calculo de la Frecuencia de excedencia en Full Spectrum LOCA mediante metodologia ISA

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Magan, J. J.; Queral Salazar, C.; Sanchez Perea, M.

    2012-07-01

    In this application LOCA sequences was taken into account the uncertainty in the size of rupture and the operator action times as cooling and depressurization through steam generators. The simulations were performed using the tool SCAIS, dynamically coupled with MAAP code.

  18. Influence of the outlet air temperature on the thermohydraulic behaviour of air coolers

    Directory of Open Access Journals (Sweden)

    Đorđević Emila M.

    2003-01-01

    Full Text Available The determination of the optimal process conditions for the operation of air coolers demands a detailed analysis of their thermohydraulic behaviour on the one hand, and the estimation of the operating costs, on the other. One of the main parameters of the thermohydraulic behaviour of this type of equipment, is the outlet air temperature. The influence of the outlet air temperature on the performance of air coolers (heat transfer coefficient overall heat transfer coefficient, required surface area for heat transfer air-side pressure drop, fan power consumption and sound pressure level was investigated in this study. All the computations, using AirCooler software [1], were applied to cooling of the process fluid and the condensation of a multicomponent vapour mixture on two industrial devices of known geometries.

  19. DEGB LOCA ECS power limit recommendation for the K-14. 1 subcycle

    Energy Technology Data Exchange (ETDEWEB)

    Smith, F.G. III; Aleman, S.E.

    1991-04-01

    This report documents assembly deposited power limits and the corresponding effluent temperature limits recommended for operating the K-14.1 subcycle to ensure sufficient cooling of reactor assemblies during the ECS phase of a Double Ended Guillotine Break (DEGSS) Loss of Coolant Accident (LOCA). The ECS LOCA effluent temperature limits are computed for each flowzone of the K-14.1 charge. The recommended overall DEGB LOCA ECS power limit is 1515 MW or about 63.1% of the historical full reactor power limit (assumed to be 2400-MW) for Mark 22 assemblies. The design basis accident is a break in the plenum inlet line where the AC pump motors not tripped.

  20. DEGB LOCA ECS power limit recommendation for the K-14.1 subcycle. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Smith, F.G. III; Aleman, S.E.

    1991-04-01

    This report documents assembly deposited power limits and the corresponding effluent temperature limits recommended for operating the K-14.1 subcycle to ensure sufficient cooling of reactor assemblies during the ECS phase of a Double Ended Guillotine Break (DEGSS) Loss of Coolant Accident (LOCA). The ECS LOCA effluent temperature limits are computed for each flowzone of the K-14.1 charge. The recommended overall DEGB LOCA ECS power limit is 1515 MW or about 63.1% of the historical full reactor power limit (assumed to be 2400-MW) for Mark 22 assemblies. The design basis accident is a break in the plenum inlet line where the AC pump motors not tripped.

  1. Validation of accelerated ageing of Thales rotary Stirling cryocoolers for the estimation of MTTF

    Science.gov (United States)

    Seguineau, C.,; Cauquil, J.-M.; Martin, J.-Y.; Benschop, T.

    2016-05-01

    The cooled IR detectors are used in a wide range of applications. Most of the time, the cryocoolers are one of the components dimensioning the lifetime of the system. The current market needs tend to reliability figures higher than 15,000hrs in "standard conditions". Field returns are hardly useable mostly because of the uncertain environmental conditions of use, or the differences in user profiles. A previous paper explains how Thales Cryogenics has developed an approach based on accelerated ageing and statistical analysis [1]. The aim of the current paper is to compare results obtained on accelerated ageing on one side, and on the other side, specific field returns where the conditions of use are well known. The comparison between prediction and effective failure rate is discussed. Moreover, a specific focus is done on how some new applications of cryocoolers (continuous operation at a specific temperature) can increase the MTTF. Some assumptions are also exposed on how the failure modes, effects and criticality analysis evolves for continuous operation at a specific temperature and compared to experimental data.

  2. Reliability improvements on Thales RM2 rotary Stirling coolers: analysis and methodology

    Science.gov (United States)

    Cauquil, J. M.; Seguineau, C.; Martin, J.-Y.; Benschop, T.

    2016-05-01

    The cooled IR detectors are used in a wide range of applications. Most of the time, the cryocoolers are one of the components dimensioning the lifetime of the system. The cooler reliability is thus one of its most important parameters. This parameter has to increase to answer market needs. To do this, the data for identifying the weakest element determining cooler reliability has to be collected. Yet, data collection based on field are hardly usable due to lack of informations. A method for identifying the improvement in reliability has then to be set up which can be used even without field return. This paper will describe the method followed by Thales Cryogénie SAS to reach such a result. First, a database was built from extensive expertizes of RM2 failures occurring in accelerate ageing. Failure modes have then been identified and corrective actions achieved. Besides this, a hierarchical organization of the functions of the cooler has been done with regard to the potential increase of its efficiency. Specific changes have been introduced on the functions most likely to impact efficiency. The link between efficiency and reliability will be described in this paper. The work on the two axes - weak spots for cooler reliability and efficiency - permitted us to increase in a drastic way the MTTF of the RM2 cooler. Huge improvements in RM2 reliability are actually proven by both field return and reliability monitoring. These figures will be discussed in the paper.

  3. Development and preliminary validation of a steam generator 3D thermohydraulics analysis code STAF

    Energy Technology Data Exchange (ETDEWEB)

    Cong, Tenglong [School of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, WI 53706 (United States); Zhang, Rui; Tian, Wenxi; Qiu, Suizheng [School of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Su, G.H., E-mail: ghsu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China)

    2016-03-15

    Highlights: • A steam generator thermohydraulics analysis code STAF is developed based on FLUENT. • Heat transfer from primary to secondary side is calculated during iteration in STAF. • Localized flow characteristics in steam generator can be obtained by STAF. • STAF is validated by simulating the AP 1000 steam generator. • STAF is validated by FRIGG test. - Abstract: Porous media model in Fluent code, coupled with two-phase mixture flow model, resistance model of tubes and heat transfer model through tubes, is employed to develop a steam generator thermohydraulics analysis code STAF (Steam generator Thermohydraulics Analysis code based on Fluent). In this code, the heat transfer from primary to secondary side is calculated three-dimensionally during iteration. The localized velocity, temperature, enthalpy, quality and void fraction in steam generator can be obtained by this code. STAF is validated in two ways. First, STAF is used to calculate the thermal-hydraulic parameters in steam generator of AP 1000. The calculated results are compared with designed values to prove that the coupled heat transfer calculation in STAF is accurate. Second, STAF is employed to simulate the FRIGG test to validate the localized parameter calculation performance by comparing the calculated localized void fraction with test values.

  4. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  5. Assessing the impact of the dispersion of fuel in case of LOCA; Evaluacion del impacto de la dispersion de combustible en caso de LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Concejal, A.; Garcia Sedano, P. J.; Crespo, A.

    2013-07-01

    Recent studies conducted in Halden and Studsvik have indicated the possibility of obtaining highly fragmented fuel with relatively low temperatures (700 degree centigrade) and high burned (70 MWd / kgU). In case of accident loss of coolant (LOCA), the expulsion may occur outside the pod fuel fragments, which can affect the coolability, cause channel blockade and therefore an increase in the maximum temperature of sheath.

  6. Atmospheric Deposition And MediterraneAN sea water productiviTy (Thales - ADAMANT) An overview

    Science.gov (United States)

    Christodoulaki, Sylvia; Petihakis, George; Triantafyllou, George; Pitta, Paraskevi; Papadimitriou, Vassileios; Tsiaras, Konstantinos; Mihalopoulos, Nikolaos; Kanakidou, Maria

    2015-04-01

    In the marine environment the salinity and biological pumps sequester atmospheric carbon dioxide. The biological pump is directly related to marine primary production which is controlled by nutrient availability mainly of iron, nitrogen and phosphorus. The Mediterranean Sea, especially the eastern basin is one of the most oligotrophic seas. The nitrogen (N) to phosphorus (P) ratio is unusually high, especially in the eastern basin (28:1) and primary production is limited by phosphorus availability. ADAMANT project contributes to new knowledge into how nutrients enter the marine environment through atmospheric deposition, how they are assimilated by organisms and how this influences carbon and nutrient fluxes. Experimental work has been combined with atmospheric and marine models. Important knowledge is obtained on nutrients deposition through mesocosm experiments on their uptake by the marine systems and their effects on the marine carbon cycle and food chain. Kinetic parameters of adsorption of acidic and organic volatile compounds in atmospheric samples of dust and marine salts are estimated in conjunction with solubility of N and P in mixtures contained in dust. Atmospheric and oceanographic models are coupled to create a system that is able to holistically simulate the effects of atmospheric deposition on the marine environment over time, beginning from the pre-industrial era until the future years (hind cast, present and forecast simulations). This research has been co-financed by the European Union (European Social Fund) and Greek national funds through the Operational Program "Education and Lifelong Learning" of the National Strategic Reference Framework - Research Funding Program: THALES, Investing in knowledge society through European Social Fund.

  7. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  8. A new species of Parategastes Sars, 1904 from the Thale Noi Lake, southern Thailand (Copepoda, Harpacticoida, Tegastidae

    Directory of Open Access Journals (Sweden)

    Thanida Saetang

    2015-09-01

    Full Text Available Parategastes pholpunthini sp. n. is described and illustrated based on material collected in the Thale Noi Lake, Phatthalung province, southern Thailand. This species can be distinguished from its congeners by the number segments of female antennule, the lengths of rami and basis of P1, the shape of middle inner seta of P4 exp-3, shape of P5, and relative lengths of spine at apically of baseoendopod of P5. The differences among Parategastes species are pointed out and they are compared with the new species. An identification key to species of the genus Parategastes are proposed.

  9. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J. [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs.

  10. Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jie; Wu, Yingwei, E-mail: wyw810@mail.xjtu.edu.cn; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-03-15

    Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li{sub 4}SiO{sub 4} lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition.

  11. Effect of Parametric Uncertainties, Variations, and Tolerances on Thermohydraulic Performance of Flat Plate Solar Air Heater

    Directory of Open Access Journals (Sweden)

    Rajendra Karwa

    2014-01-01

    Full Text Available The paper presents results of an analysis carried out using a mathematical model to find the effect of the uncertainties, variations, and tolerances in design and ambient parameters on the thermohydraulic performance of flat plate solar air heater. Analysis shows that, for the range of flow rates considered, a duct height of 10 mm is preferred from the thermohydraulic consideration. The thermal efficiency changes by about 2.6% on variation in the wind heat transfer coefficient, ±5 K variation in sky temperature affects the efficiency by about ±1.3%, and solar insolation variation from 500 to 1000 Wm−2 affects the efficiency by about −1.5 to 1.3% at the lowest flow rate of 0.01 kgs−1 m−2 of the absorber plate with black paint. In general, these effects reduce with increase in flow rate and are lower for collector with selective coating on the absorber plate surface. The tolerances in the duct height and absorber plate emissivity should be small while positive tolerance of 3° in the collector slope for winter operation and ±3° for year round operation, and a positive tolerance for the gap between the absorber plate and glass cover at nominal value of 40 mm are recommended.

  12. Validation and verification of the MTR{sub P}C thermohydraulic package

    Energy Technology Data Exchange (ETDEWEB)

    Doval, Alicia [INVAP S.E., Bariloche, Rio Negro (Argentina). Nuclear Engineering Dept.]. E-mail: doval@invap.com.ar

    1998-07-01

    The MTR{sub P}C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

  13. Heat flux: thermohydraulic investigation of solar air heaters used in agro-industrial applications

    Science.gov (United States)

    Rahmati Aidinlou, H.; Nikbakht, A. M.

    2017-03-01

    A new design of solar air heater simulator is presented to comply with the extensive applications inagro-industry. A wise installation of increased heat transfer surface area provided uniform and efficient heat diffusion over the duct. Nusselt number and friction factor have been investigated based on the constant roughness parameters such as relative roughness height (e/D), relative roughness pitch (P/e), angle of attack (α) and aspect ratio with Reynolds numbers ranging from 5000 to 19,000 in the fully developed region. Heat fluxes of 800, 900 and 1000 Wm-2 were provided. The enhancement in friction factor is observed to be 3.1656, 3.47 and 3.0856 times, and for the Nusselt number either, augmentation is calculated to be 1.4437, 1.4963 and 1.535 times, respectively, over the smooth duct for 800, 900 and 1000 Wm-2 heat fluxes. Thermohydraulic performance is plotted versus the Reynolds number based on the aforementioned roughness parameters at varying heat fluxes. The results show up that thermohydraulic performance is found to be maximum for 1000 Wm-2 at the average Reynolds number of 5151. Based on the results, we can verify that the introduced solar simulator can help analyzing and developing solar collector installations at the simulated heat fluxes.

  14. A novel thermo-hydraulic coupling model to investigate the crater formation in electrical discharge machining

    Science.gov (United States)

    Tang, Jiajing; Yang, Xiaodong

    2017-09-01

    A novel thermo-hydraulic coupling model was proposed in this study to investigate the crater formation in electrical discharge machining (EDM). The temperature distribution of workpiece materials was included, and the crater formation process was explained from the perspective of hydrodynamic characteristics of the molten region. To better track the morphology of the crater and the movement of debris, the level-set method was introduced in this study. Simulation results showed that the crater appears shortly after the ignition of the discharge, and the molten material is removed by vaporizing in the initial stage, then by splashing at the following time. The driving force for the detachment of debris in the splashing removal stage comes from the extremely large pressure difference in the upper part of the molten region, and the morphology of the crater is also influenced by the shearing flow of molten material. It was found that the removal ratio of molten material is only about 7.63% under the studied conditions, leaving most to form the re-solidification layer on the surface of the crater. The size of the crater reaches the maximum at the end of discharge duration then experiences a slight reduction because of the reflux of molten material after the discharge. The results of single pulse discharge experiments showed that the morphologies and sizes between the simulation crater and actual crater are good at agreement, verifying the feasibility of the proposed thermo-hydraulic coupling model in explaining the mechanisms of crater formation in EDM.

  15. Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Oezdemir, Erdal; Moon, Kang Hoon; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Kim, Yongdeog [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES.

  16. Política habitacional e locação social em Salvador

    Directory of Open Access Journals (Sweden)

    Nelson Baltrusis

    Full Text Available Este artigo tem como objetivo analisar o mercado imobiliário de locação em Salvador. Num primeiro momento, caracterizaremos o problema habitacional em Salvador, para o que nos apoiaremos nas diretrizes e ações previstas no Plano Municipal de Habitação de Interesse Social (PMHIS. Em seguida, trataremos das políticas implantadas pelos governos do estado e federal, destacando a experiência do Programa de Arrendamento Residencial (PAR e incorporando algumas considerações sobre o Programa Minha Casa, Minha Vida. Também será abordada a questão do mercado de locação em Salvador a partir do perfil de moradores e da dinâmica do mercado.

  17. Assessment of CONTAIN and MELCOR for performing LOCA and LOVA analyses in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J.; Hagrman, D.L.; Gaeta, M.J.; Petti, D.A.

    1994-09-01

    This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe accident thermal hydraulic analyses. This code will require the least modification to be appropriate for calculating thermal hydraulic behavior in ITER relevant conditions that include vacuum, cryogenics, ITER temperatures, and the presence of a liquid metal test module. The assessment of the aerosol transport models in these codes concludes that several modifications would have to be made to CONTAIN and/or MELCOR to make them applicable to the aerosol transport part of severe accident analysis in ITER.

  18. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F.; Gauthier, G.; Carlin, F. [and others

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40{degrees}C or 70{degrees}C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased.

  19. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko, M.; Antti, D.; Hanna, R.; Timo, V. [VTT Processes, (Finland)

    2004-07-01

    TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable part in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but the rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. (authors)

  20. Change detection and identification of land potential for planting Krajood (Lepironia articulata in Thale Noi, Southern Thailand

    Directory of Open Access Journals (Sweden)

    Jitnapa Maeaid

    2012-07-01

    Full Text Available Lepironia articulata, commonly called grey sedge or krajood, can be transformed into various products to generateextra income for local families in the southern part of Thailand. In recent years, the amount of Lepironia articulata used asraw material has decreased and does not currently meet the demand for the resource. Appropriate areas where naturalresources and the environment can be restored and the abundance of natural produce can be increased must be sought.Therefore, this research considered the opportunity to identify appropriate areas for planting Lepironia articulata. Geographicinformation system (GIS and remote sensing were integrated to map land use changes in 1990, 1998 and 2006 in theThale Noi area. The study found that from 1990-1998, emergent aquatic areas increased by 16.18 square kilometers, the areaof swamp forests increased by 15.33 square kilometers, the area of rice paddies decreased by 0.80 square kilometers, and thearea of mixed orchards increased by approximately 0.32 square kilometers. From 1998-2006, the area of swamp forestsincreased by 1.9 square kilometers, but emergent aquatic areas decreased by 1.23 square kilometers. The area of rubberplantations increased by 0.63 square kilometers, and the area of rice paddies decreased by 0.69 square kilometers. This studyaimed to define land potential for Krajood (Lepironia articulata cultivation in the Thale Noi area by considering five factors:land use, distance from water sources, slope, soil characteristics, and soil drainage. The study found that the areas of highpotential for planting Lepironia articulata were wetlands and near water sources, covering a total area of 5.54 square kilometers.The areas with moderate potential were swamp forests and rice paddies, covering a total area of 4.27 square kilometres.GIS and remote sensing were found to be very useful for identifying land use changes and potential areas for plantingLepironia articulata.

  1. Design, construction and evaluation of solar flat-plate collector simulator based on the thermohydraulic coefficient

    Directory of Open Access Journals (Sweden)

    H Rahmati Aidinlou

    2017-05-01

    Full Text Available Introduction Increasing the area of absorber plate between the flowed air through the duct can be accomplished by corrugating the absorber plate or by using the artificial roughness underside of the absorber plate as the commercial methods for enhancing the thermohydraulic performance of the flat plate solar air heaters. Evaluation of this requires the construction of separated solar air heater which is costly and time consuming. The constructed solar flat-plate collector simulator can be a sufficient solution for obtaining the heat transfer and thermodynamic parameters for evaluating the absorber plate. The inclined broken roughness was chosen as the optimum roughness which is surrounded by three aluminum smooth walls. Materials and Methods The duct for both smooth and roughened plate have been constructed based on the ASHRAE 93-2010 standard. In order to achieve a fully thermal and hydraulic developed flow, the plenum is constructed. The centrifugal fan is considered by applying the required air volume at the pressure drop obtained by the duct, plenum and the orifice meter. The TSI velocity-meter 8355 is used to measure the velocity of air crossing through the pipe connected to the centrifugal fan. The micro manometer Kimo CPE310-s with the resolution of 0.1 Pa is used to measure the pressure drop across the test section of the smooth and roughened duct. The LM35 sensors are used to measure the absorber plate and air temperature through the test section. Obtained parameters are used to calculate the Nusselt number and friction factor across the test section for smooth and roughened absorber plate. The Nusselt number and friction factor parameters which is obtained for smooth absorber plate based on experimental set-up, is compared with Dittus-Bolter and Blasius equations, respectively, for validating the simulator. By calculating the Nusselt number and friction factor, Stanton number is obtained based on the equation (6, and thermohydraulic

  2. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  3. R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron Simon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tu, Lei [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize

  4. Optimum control parameters and long-term productivity of geothermal reservoirs using coupled thermo-hydraulic process modelling

    OpenAIRE

    Aliyu, Musa D.; Chen, Hua-Peng

    2017-01-01

    Knowing the long-term performance of geothermal energy extraction is crucial to decision-makers and reservoir engineers for optimal management and sustainable utilisation. This article presents a three dimensional, numerical model of coupled thermo-hydraulic processes, in a deep heterogeneous geothermal reservoir overlain and underlain by impermeable layers, with discrete fracture. The finite element method is employed in modelling the reservoir, after conducting a verification study to test ...

  5. Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.

  6. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Diamond, D. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-08-30

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  7. IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS

    Directory of Open Access Journals (Sweden)

    DONG HYUN LEE

    2014-08-01

    Full Text Available Probabilistic Safety Assessment (PSA has been widely used to estimate the overall safety of nuclear power plants (NPP and it provides base information for risk informed application (RIA and risk informed regulation (RIR. For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

  8. Water volume available for ECCS sump recirculation mode following a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Riekert, T. [TUV NORD SysTec (Germany); Rebohm, H. [TUV NORD EnSys Hannover (Germany); Huber, J. [TUV SUD IS (Germany); Brandes, F. [TUV SUD ET (Germany)

    2006-07-01

    In this paper we describe the reviews performed in Germany on the water level in the containment sump after a LOCA and the derived actions. Our view on the issue is from the perspective of the independent safety experts - i.e. TUV SUD Industrie Service (TUV SUD IS), TUV SUD Energietechnik GmbH Baden-Wuerttemberg (TUV SUD ET), TUV NORD EnSys Hannover and TUV NORD SysTec -, which reviewed the analyses of the utilities on behalf of the responsible supervising authorities. Between these expert organizations information were exchanged via the steering committee on nuclear technology of the association of the TUVs (VdTUV). In our paper we describe the analyses on the two safety issues relevant in the connection with the water level in the containment sump: the necessary minimum coverage of suction pipes to avoid inadmissible entrainment of air and the water retention inside the containment after a LOCA. Our description concentrates on PWRs because of the more complex conditions in comparison to BWRs. In conclusion it can be stated that due to the thorough evaluation of operating experience, optimization measures could be derived. In addition, the analyses served the purpose of know-how maintenance. (authors)

  9. Heat transfer processes during intermediate and large break loss-of-coolant accidents (LOCAs)

    Energy Technology Data Exchange (ETDEWEB)

    Vojtek, I

    1986-09-01

    The general purpose of this project was the investigation of the heat transfer regimes during the high pressure portion of blowdown. The main attention has been focussed on the evaluation of those phenomena which are most important in reactor safety, such as maximum and minimum critical heat flux and forced convection film boiling heat transfer. The experimental results of the 25-rod bundle blowdown heat transfer tests, which were performed at the KWU heat transfer test facility in Karlstein, were used as a database for the verification of different correlations which are used or were developed for the analysis of reactor safety problems. The computer code BRUDI-VA was used for the calculation of local values of important thermohydraulic parameters in the bundle.

  10. Response of Compacted Bentonites to Thermal and Thermo-Hydraulic Loadings at High Temperatures

    Directory of Open Access Journals (Sweden)

    Snehasis Tripathy

    2017-07-01

    Full Text Available The final disposal of high-level nuclear waste in many countries is preferred to be in deep geological repositories. Compacted bentonites are proposed for use as the buffer surrounding the waste canisters which may be subjected to both thermal and hydraulic loadings. A significant increase in the temperature is anticipated within the buffer, particularly during the early phase of the repository lifetime. In this study, several non-isothermal and non-isothermal hydraulic tests were carried on compacted MX80 bentonite. Compacted bentonite specimens (water content = 15.2%, dry density = 1.65 Mg/m3 were subjected to a temperature of either 85 or 150 °C at one end, whereas the temperature at the opposite end was maintained at 25 °C. During the non-isothermal hydraulic tests, water was supplied from the opposite end of the heat source. The temperature and relative humidity were monitored along predetermined depths of the specimens. The profiles of water content, dry density, and degree of saturation were established after termination of the tests. The test results showed that thermal gradients caused redistribution of the water content, whereas thermo-hydraulic gradients caused both redistribution and an increase in the water content within compacted bentonites, both leading to development of axial stress of various magnitudes. The applied water injection pressures (5 and 600 kPa and temperature gradients appeared to have very minimal impact on the magnitude of axial stress developed. The thickness of thermal insulation layer surrounding the testing devices was found to influence the temperature and relative humidity profiles thereby impacting the redistribution of water content within compacted bentonites. Under the influence of both the applied thermal and thermo-hydraulic gradients, the dry density of the bentonite specimens increased near the heat source, whereas it decreased at the opposite end. The test results emphasized the influence of

  11. Política habitacional no Brasil e Programa de Locação Social paulistano

    Directory of Open Access Journals (Sweden)

    Camila D'Ottaviano

    Full Text Available O artigo apresenta uma breve analise da política habitacional brasileira, tendo como estudo de caso específico o Programa de Locação Social instituído no município de São Paulo em 2002. O Programa de Locação Social paulistano é o principal programa habitacional alicerçado na locação implantado no Brasil recentemente. Criado pela Resolução nº 23 do Conselho do Fundo Municipal de Habitação, visava atender à demanda das famílias com renda de até três salários mínimos, excluídas até então dos programas existentes de financiamento habitacional. Após 10 anos de sua implantação, as principais limitações apontadas são a falta de acompanhamento social da demanda atendida e a falta de acompanhamento na gestão das áreas condominiais. Outra questão ainda não resolvida, no âmbito do programa de locação social, é o tempo máximo de permanência das famílias nos empreendimentos, bem como o atendimento a uma demanda muito superior à capacidade do Programa.

  12. Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yurko, Joseph P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swindlehurst, Gregg [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

  13. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  14. Thermohydraulic behavior of the liquid metal target of a spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Y.

    1996-06-01

    The author presents work done on three main problems. (1) Natural circulation in double coaxial cylindircal container: The thermohydraulic behaviour of the liquid metal target of the spallation neutron source at PSI has been investigated. The configuration is a natural-circulation loop in a concentric double-tube-type container. The results show that the natural-circulation loop concept is valid for the design phase of the target construction, and the current specified design criteria will be fulfilled with the proposed parameter values. (2) Flow around the window: Water experiments were performed for geometry optimisation of the window shape of the SINQ container for avoiding generating recirculation zones at peripheral area and the optimal cooling of the central part of the beam entrance window. Flow visualisation technique was mainly used for various window shapes, gap distance between the window and the guide tube edge. (3) Flow in window cooling channels: Flows in narrow gaps of cooling channels of two different types of windows were studied by flow visualisation techniques. One type is a slightly curved round cooling channel and the other is hemispherical shape, both of which have only 2 mm gap distance and the water inlet is located on one side and flows out from the opposite side. In both cases, the central part of the flow area has lower velocity than peripheral area.

  15. Effect of bundle size on cladding deformation in LOCA simulation tests. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, R.H.; Crowley, J.L.; Longest, A.W.

    1982-01-01

    Two LOCA simulation tests were conducted to investigate the effects of temperature uniformity and radial restraint boundary conditions on Zircaloy cladding deformation. In one of the tests (B-5), boundary conditions typical of a large array were imposed on an inner 4 x 4 square array by two concentric rings of interacting guard fuel pin simulators. In the other test (B-3), the boundary conditions were imposed on a 4 x 4 square array by a non-interacting heated shroud. Test parameters conducive to large deformation were selected in order to favor rod-to-rod interactions. The tests showed that rod-to-rod interactions play an important role in the deformation process.

  16. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  17. Estimation of Leak Flow Rate during Post-LOCA Using Cascaded Fuzzy Neural Networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    In this study, important parameters such as the break position, size, and leak flow rate of loss of coolant accidents (LOCAs), provide operators with essential information for recovering the cooling capability of the nuclear reactor core, for preventing the reactor core from melting down, and for managing severe accidents effectively. Leak flow rate should consist of break size, differential pressure, temperature, and so on (where differential pressure means difference between internal and external reactor vessel pressure). The leak flow rate is strongly dependent on the break size and the differential pressure, but the break size is not measured and the integrity of pressure sensors is not assured in severe circumstances. In this paper, a cascaded fuzzy neural network (CFNN) model is appropriately proposed to estimate the leak flow rate out of break, which has a direct impact on the important times (time approaching the core exit temperature that exceeds 1200 .deg. F, core uncover time, reactor vessel failure time, etc.). The CFNN is a data-based model, it requires data to develop and verify itself. Because few actual severe accident data exist, it is essential to obtain the data required in the proposed model using numerical simulations. In this study, a CFNN model was developed to predict the leak flow rate before proceeding to severe LOCAs. The simulations showed that the developed CFNN model accurately predicted the leak flow rate with less error than 0.5%. The CFNN model is much better than FNN model under the same conditions, such as the same fuzzy rules. At the result of comparison, the RMS errors of the CFNN model were reduced by approximately 82 ~ 97% of those of the FNN model.

  18. H2S AND NO SIGNALING INTERACTIONS IN THALE CRESS (ARABIDOPSIS THALIANA L. AND PEPPER (CAPSICUM ANNUUM L. LEAVES

    Directory of Open Access Journals (Sweden)

    Miroslav Lisjak

    2012-06-01

    Full Text Available This research comprehends a set of experiments with several thale cress (Arabidopsis thaliana L. and pepper (Capsicum annuum L. genotypes in controlled conditions using growth chambers, with the aim of determining the physiological role of hydrogen sulfide (H2S in plants, as well as its potential effect as a signaling compound, particularly in potential interaction with nitric oxide (NO signaling pathways. Special emphasis was focused on stomatal mechanisms and signaling in their opening and closing. Moreover, the effect of treatment of pepper plants with H2S was investigated in salt stress conditions. It was established that the applied H2S donors, NaHS and GYY4137, inhibit stomata closing in both plant species through the reduction of NO accumulation in stomata, which was proven to occur in SNP or ABA treatment. The effects of NO and H2S were opposite those in pepper plants response to salt stress as well, with increased antioxidative activity in leaf obtained after H2S treatments, and with NaHS in particular. In addition, GYY4137 could be considered as a convenient H2S donor for research into H2S functions in plants. The results point out the interactions of H2S and NO in plant cell signaling in both normal and salt stress conditions. Further research of this type should uncover H2S functions in plant metabolism more precisely, especially considering the potential practical value of this knowledge for plant stress resistance improvement and their productivity enhancement.

  19. La construccion de "la loca" en dos novellas chilenas: El lugar sin limites de Jose Donoso y Tengo miedo torero de Pedro Lemebel

    National Research Council Canada - National Science Library

    Lopez Morales, Berta

    2011-01-01

    Este articulo intenta mostrar como se construye la identidad sexual y social del personaje 'la loca"en dos novelas chilenas, El lugar sin limites de Jose Donoso y Tengo miedo torero de Pedro Lemebel...

  20. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  1. Computational fluid dynamic model for thermohydraulic calculation for the steady-state of the real scale HTR-1

    Energy Technology Data Exchange (ETDEWEB)

    Gamez, Abel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy Y.; Gonzalez, Daniel; Garcia, Carlos, E-mail: agamezgmf@gmail.com, E-mail: leored1984@gmail.com, E-mail: jrosales@instec.cu, E-mail: lcastro@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgr@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Oliveira, Carlos B. de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)

    2015-07-01

    The high temperature gas cooled reactor (HTGR) is one of candidates of next generation of nuclear reactor according to IAEA report 2013. Evaluation of thermohydraulic performance and an experimental comparison results were proposed to the international research community. In this article, the tree dimensional CFD thermohydraulic modelation of steady state of HTR-10 modular reactor, using ANSYS CFX v14.0, has been done. Code-to-code and Code-to-experiment benchmark analyses, related to the testing program of the HTR-10 plant including steady state temperature distribution with the reactor at full power, were developed. The 3D real scale representation of reflector zone and fluid path flow inner and outer reflector blocks and cold helium cavity were carried out. The porous medium model was used to simulate the core zone in the reactor. The power distribution of the initial core published by IAEA-TECDOC-1694 obtained by Chief Scientific Investigators (CSIs) from China was used as heat sources in the core zone. (author)

  2. A Demonstration of Advanced Safety Analysis Tools and Methods Applied to Large Break LOCA and Fuel Analysis for PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Laboratory; Smith, Curtis Lee [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2016-03-01

    The U.S. Nuclear Regulatory Commission (NRC) is currently proposing a rulemaking designated as 10 CFR 50.46c to revise the loss-of-coolant accident (LOCA)/emergency core cooling system acceptance criteria to include the effects of higher burnup on fuel/cladding performance. We propose a demonstration problem of a representative four-loop PWR plant to study the impact of this new rule in the US nuclear fleet. Within the scope of evaluation for the 10 CFR 50.46c rule, aspects of safety, operations, and economics are considered in the industry application demonstration presented in this paper. An advanced safety analysis approach is used, by integrating the probabilistic element with deterministic methods for LOCA analysis, a novel approach to solving these types of multi-physics, multi-scale problems.

  3. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko Miettinen; Timo Vanttola; Hanna Raety; Antti Daavittila [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

    2005-07-01

    Full text of publication follows: TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four-equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one-dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five-equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. In general questions could be raised, what advantages are seen with the new internal coupling in comparison with the earlier realised parallel coupling, and which advantages may be seen in building the realtor physical model on the basis of the old code, developed since 1970's. The internal coupling allows

  4. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

    OpenAIRE

    Yuquan, Li; Botao, Hao; Jia, Zhong; Nan, Wang

    2017-01-01

    The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although so...

  5. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken.

  6. Thermo-Hydraulic Analyses Associated with the Design of JT-60SA TF Coils Development and Validation of the Tacos/texto Tool

    Science.gov (United States)

    Lacroix, B.; Portafaix, C.; Hertout, P.; Nicollet, S.; Zani, L.; Barabaschi, P.; Villari, R.

    2010-04-01

    In the framework of the JT-60SA project, TF coils design activities have led to the development of TEXTO (Thermo-hydraulic EXtended TOol), a dedicated simulation tool centred on the GANDALF code for calculating the temperature margin (ΔTma) central criterion. From a first version providing conservative results, this tool has been upgraded to a pseudo 3D model by integrating a 1D thermo-hydraulic approach in ANSYS, leading to an additional and independent module named TACOS (Thermo-hydraulic Ansys COmputation Semi 3D). By providing He temperature in conductors, TACOS is well adapted for evaluating the impact of TF coils design choices in the framework of cost and feasibility optimization. TACOS also provides more accurate values of transverse heat flux from case to conductors, which can then be injected into TEXTO and allow the calculation of the temperature margin, thanks to the functionalities of GANDALF. Several validation calculations have been performed for the reference operation scenario, by comparison with GANDALF for validating the simplified thermo-hydraulic method of TACOS and by comparison with VINCENTA for an overall validation of the TACOS/TEXTO tool. Results have shown a good consistency between the different models, with a conductor temperature dispersion around 0.05 K in the critical zone.

  7. Fission product transport analysis: Task 2. Quarterly progress report, April--June 1977. [PWR; BWR; primary system reflooding following LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Gieseke, J.A.; Jordan, H.; Baybutt, P.; Wooton, R.O.; Denning, R.S.

    1977-09-30

    Continuing activities associated with modeling the transport and deposition of fission products within PWR and BWR primary systems during the reflood time period following a terminated LOCA are reported. The original scope of the overall project has been expanded to include consideration of conditions consistent with those leading to postulated core meltdown situations. Initial tasks on this continuation study involve limited improvements to the TRAP codes and performance of sensitivity analyses for terminated LOCA's plus the evaluation of thermal-hydraulic conditions for use in evaluating fission product transport and deposition in postulated meltdown situations. Beyond these initial tasks, emphasis will be concentrated solely on meltdown analyses. Major efforts within the past quarter have included completion in final form of the informal interim report on analyses for terminated LOCA conditions, initiation of improvement for the TRAP codes to include additional fission product deposition mechanisms, consideration of suitable methods for performing the sensitivity analyses, specification of ranges for variables to be covered in the sensitivity analyses, and initiation of efforts to specify flow conditions to be assumed in future development of analysis procedures applicable to meltdown conditions.

  8. Assessment of Environmental Qualification Practices and Condition Monitoring Techniques for Low-Voltage Electric Cables: LOCA Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Lofaro, R. [Brookhaven National Lab. (BNL), Upton, NY (United States); Grove, E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Villaran, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Soo, P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hsu, F. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2001-02-01

    This report documents the results of a research program addressing issues related to the qualification process for low-voltage instrumentation and control (I&C) electric cables used in commercial nuclear power plants. Three commonly used types of I&C cable were tested: Cross-Linked Polyethylene (XLPE) insulation with a Neoprene® jacket, Ethylene Propylene Rubber (EPR) insulation with an unbonded Hypalon® jacket, and EPR with a bonded Hypalon® jacket. Each cable type received accelerated aging to simulate 20, 40, and 60 years of qualified life. In addition, naturally aged cables of the same types were obtained from decommissioned nuclear power plants and tested. The cables were subjected to simulated loss-of-coolant-accident (LOCA) conditions, which included the sequential application of LOCA radiation followed by exposure to steam at high temperature and pressure, as well as to chemical spray. Periodic condition monitoring (CM) was performed using nine different techniques to obtain data on the condition of the cable, as well as to evaluate the effectiveness of those CM techniques for in situ monitoring of cables. Volume 1 of this report presents the results of the LOCA tests, and Volume 2 discusses the results of the condition monitoring tests.

  9. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    Energy Technology Data Exchange (ETDEWEB)

    Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

    2015-11-15

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  10. RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

    Highlights: • RELAP5/SCDAPSIM model of AP1000 has been developed. • Analysis involves a LBLOCA (double ended guillotine break) study in cold leg. • Results are compared with those of WCOBRA–TRAC and TRACE. • Concluded that PCT does not violate the safety criteria of 1477 K. - Abstract: The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved

  11. Oxidation investigation of cladding specimens for regular and accident tolerant fuel rods under LOCA conditions

    Science.gov (United States)

    Bazyuk, S. S.; Deryabin, I. A.; Kiselev, D. S.; Kuzma-Kichta, Yu A.; Mokrushin, A. A.; Parshin, N. Ya; Popov, E. B.; Soldatkin, D. M.

    2017-11-01

    The high-temperature oxidation tests were carried out for the regular fuel rod claddings specimens made of sponge-based zirconium alloy (E110G) and for the accident tolerant fuel (ATF) ones – pure vacuum melted molybdenum (VCPM) and niobium alloy (Nb-1%Zr). The tests were carried out under the ambient pressure p ∼ 0.1 MPa in pure water steam. The experimental data on the oxidation characteristics were obtained for E110G specimens in the temperature range T = 1100 ‑ 1500 °C, that for VCPM and Nb-1%Zr are investigated under extended temperature-duration range (more than 1 hour). The thermal effects of molybdenum (QSMR) and niobium (QSNR) interactions with steam were defined and the derived oxidation rate constants for refractory metals were compared with the known ones. Based on the computations performed with PARAM-TG code the high-temperature oxidation characteristics of model fuel assemblies of large-scale facilities under LOCA conditions with regular and ATF claddings were compared. It was shown that Zr-steam interaction of fuel rod cladding (QSZR) is more intensive compared with VCPM and Nb-1%Zr ones under investigated conditions.

  12. A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

    Directory of Open Access Journals (Sweden)

    Jae-Yong Kim

    2014-01-01

    Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

  13. New roles for astrocytes: the nightlife of an 'astrocyte'. La vida loca!

    Science.gov (United States)

    Horner, Philip J; Palmer, Theo D

    2003-11-01

    Like a newly popular nightspot, the biology of adult stem cells has emerged from obscurity to become one of the most lively new disciplines of the decade. The neurosciences have not escaped this trendy pastime and, from amid the noise and excitement, the astrocyte emerges as a beguiling companion to the adult neural stem cell. A once receding partner to neurons and oligodendrocytes, the astrocyte even takes on an alter ego of the stem cell itself (S. Goldman, this issue of TINS). Putting ego aside, the 'astrocyte' is also (and perhaps more importantly) an integral component of neural progenitor hotspots, where the craziness or 'la vida loca' of the nightlife might not be so wild when compared with our traditional understanding of the astrocyte. Here, astrocytes contribute to the instructive confluence of location, atmosphere and cellular neighbors that define the daily 'vida local' or everyday local life of an adult stem cell. This review discusses astrocytes as influential components in the local stem cell niche.

  14. Prediction of Reactor Vessel Water Level Using Fuzzy Neural Networks in Severe Accidents due to LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soonho; Kim, Jaehawn; Na, Mangyun [Chosun Univ., Gwangju (Korea, Republic of)

    2013-05-15

    When the initial events that may lead to the severe accident such as Loss Of Coolant Accident (LOCA) and Steam Generator Tube Rupture (SGTR) occurs at a nuclear power plant, it is most important to check the status of the plant conditions by observing the safety-related parameters such as neutron flux, pressurizer pressure, steam generator pressure and water level. In this paper, we propose a method of predicting the water level of coolant in the reactor vessel that directly affect the important events such as the exposure of the reactor core and the damage of reactor vessel by using a Fuzzy Neural Network (FNN) method. In addition, the data for verifying a proposed model was obtained by simulating the severe accident scenarios for the OPR1000 nuclear power plant using the MAAP4 code. In this paper, a prediction model was developed for predicting the reactor vessel water level using the FNN method. The proposed FNN model was verified based on the simulation data of OPR1000 by using MAAP4 code. As a result of simulation, we could see that the performance of the proposed FNN model is quite satisfactory but some large errors are observed occasionally. If the proposed FNN model is optimized by using a variety of data, it is possible to predict the reactor vessel water level exactly.

  15. Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

    Directory of Open Access Journals (Sweden)

    Lázaro Lima

    2011-12-01

    Full Text Available

    Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

  16. Design of Test Facility to Evaluate Boric Acid Precipitation Following a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong-Kwan; Song, Yong-Jae [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The U.S.NRC has identified a concern that debris associated with generic safety issue (GSI) - 191 may affect the potential precipitation of boric acid due to one or more of the following phenomena: - Reducing mass transport (i.e. mixing) between the core and the lower plenum (should debris accumulate at the core inlet) - Reduced lower plenum volume (should debris settle in the lower plenum), and, - Increased potential for boric acid precipitation (BAP) in the core (should debris accumulate in suspension in the core) To address these BAP issues, KHNP is planning to conduct validation tests by constructing a BAP test facility. This paper describes the design of test facility to evaluate BAP following a LOCA. The design of BAP test facility has been developed by KHNP. To design the test facility, test requirements and success criteria were established, and scaling analysis of power-to-volume method, Ishii-Kataoka method, and hierarchical two-tiered method were investigated. The test section is composed of two fuel assemblies with half of full of prototypic FA height. All the fuel rods are heated by the electric power supplier. The BAP tests in the presence of debris, buffering agents, and boron will be performed following the test matrix.

  17. Thermohydraulic analysis of high-Prandtl-number fluid in complex duct simulating first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Satake, Masaaki [Advanced Fusion Reactor Engineering Laboratory, Department of Quantum Science and Energy Engineering, Graduate School of Engineering Tohoku University, Aramaki-Aza-Aoba 6-6-01-2, Aoba-ku, Sendai, Miyagi 980-8579 (Japan)], E-mail: msata@karma.qse.tohoku.ac.jp; Yuki, Kazuhisa [Advanced Fusion Reactor Engineering Laboratory, Department of Quantum Science and Energy Engineering, Graduate School of Engineering Tohoku University, Aramaki-Aza-Aoba 6-6-01-2, Aoba-ku, Sendai, Miyagi 980-8579 (Japan)], E-mail: kyuki@qse.tohoku.ac.jp; Hashizume, Hidetoshi [Advanced Fusion Reactor Engineering Laboratory, Department of Quantum Science and Energy Engineering, Graduate School of Engineering Tohoku University, Aramaki-Aza-Aoba 6-6-01-2, Aoba-ku, Sendai, Miyagi 980-8579 (Japan)], E-mail: hidetoshi.hashizume@qse.tohoku.ac.jp

    2010-04-15

    For fusion reactors, molten salt is one of the candidates for coolant materials. Molten salt is a high-Prandtl-number fluid; thus, it is necessary to enhance the heat transfer coefficient. It is proposed that rods are inserted into a duct to enhance the heat transfer coefficient. The flow field behind the rod in the duct is visualized to compare experimental data with simulation results. The trends and distributions in the numerical simulation are the same as those in the experiment, and furthermore, the magnitudes of the time and space scales in the numerical simulation are of the same order as those in the experiment. Thermohydraulic numerical analysis confirmed that the heat transfer coefficient is improved by inserting the rod when the fluid is a high-Prandtl-number fluid and the flow field is in the turbulent region. However, it is necessary for the rods to be arranged in the streamwise direction.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  19. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  20. Oxygen segregation in pre-hydrided Zircaloy-4 cladding during a simulated LOCA transient

    Directory of Open Access Journals (Sweden)

    Torres Elodie

    2017-01-01

    Full Text Available Oxygen and hydrogen distributions are key elements influencing the residual ductility of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA. During the high temperature oxidation, a complex partitioning of the alloying elements is observed. A finite-difference code for solving the oxygen diffusion equations has been developed by Institut de Radioprotection et de Sûreté Nucléaire to predict the oxygen profile within the samples. The comparison between the calculations and the experimental results in the mixed α+β region shows that the oxygen diffusion is not accurately predicted by the existing modeling. This work aims at determining the key parameters controlling the average oxygen profile within the sample in the two-phase regions at 1200 °C. High temperature steam oxidation tests interrupted by water quench were performed using pre-hydrided Zircaloy-4 samples. Experimental oxygen distribution was measured by Electron Probe Micro-Analysis (EPMA. The phase distributions within the cladding thickness, was measured using image analysis to determine the radial profile of α(O phase fraction. It is further demonstrated and experimentally checked that the α-phase fraction in these regions follows a diffusion-like radial profile. A new phase fraction modeling is then proposed in the cladding metallic part during steam oxidation. The modeling results are compared to a large set of experiments including the influence of exposure duration and hydrogen content. Another key outcome from this modeling is that oxygen average profile is straightforward derived from the proposed modeling.

  1. Prediction of Leak Flow Rate Using FNNs in Severe LOCA Circumstances

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Kim, Ju Hyun; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of); Hur, Seop; Kim, Chang Hwoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Leak flow rate is a function of break size, differential pressure ( i.e., difference between internal and external reactor vessel pressure), temperature, and so on. Specially, the leak flow rate is strongly dependent on the break size and the differential pressure, but the break size is not measured and the integrity of pressure sensors is not assured in severe circumstances. In this study, a fuzzy neural network (FNN) model is proposed to predict the leak flow rate out of break, which has a direct impact on the important times (time approaching the core exit temperature that exceeds 1200 .deg. F, core uncover time, reactor vessel failure time, etc.). Since FNN is a data-based model, it requires data to develop and verify itself. However, because actual severe accident data do not exist to the best of our knowledge, it is essential to obtain the data required in the proposed model using numerical simulations. These data were obtained by simulating severe accident scenarios for the optimized power reactor 1000 (OPR 1000) using MAAP4 code. In this study, FNN model was developed to predict the leak flow rate in severe post-LOCA circumstances.. The training data were selected from among all the acquired data using an SC method to train the proposed FNN model with more informative data. The developed FNN model predicted the leak flow rate using the time elapsed after reactor shutdown and the predicted break size, and its validity was verified in the basis of the simulation data of OPR1000 using MAAP4 code.

  2. Effect of In-core Blockage by Debris during Post-LOCA Long Term Core Cooling Phase of Kori-2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taewan [Incheon National University, Incheon (Korea, Republic of); Jin, Chang-Yong; Bang, Young-Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Generic Safety Issue (GSI) 191 concerns the degradation of heat transfer in the core during Post loss-of-coolant-accident (LOCA) long term core cooling (LTCC) phase by debris which may go through the sump strainer and could be deposited at the core inlet and fuel surface. United State Nuclear Regulatory Commission (US NRC) approved a generic and conservative methodology described in WCAP-16793-NP Rev. 2, and has made use of it for GSI-191 resolution. In Korea, as a part of periodic safety review of Kori-2, an evaluation of thermal hydraulic effect of in-core blockage by debris has carried out based on a conservative emergency core cooling system (ECCS) evaluation method (EM). This paper describes a realistic approach to evaluate the thermal hydraulic effect of in-core blockage by debris during post-LOCA LTCC of Kori-2. The MARS-KS 1.3 code has been employed for the thermal hydraulic analysis. The effect of in-core blockage by debris has been evaluated by thermal hydraulic analyses with MARS-KS. In order to evaluate the heat transfer degradation by debris deposition a conservative and realistic fuel models has been developed, respectively. The analysis indicates that the PCT during the post-LOCA LTCC phase increases due to the heat transfer degradation by debris deposition and flow reduction by in-core blockage. It is also found that the PCT increases more in hot leg break case because of a larger reduction in core flow by higher pressure drop at the core inlet.

  3. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  4. Testing to evaluate synergistic effects from LOCA environments. Test IX. Simultaneous mode; cables, splice assemblies, and electrical insulation samples

    Energy Technology Data Exchange (ETDEWEB)

    Thome, F.V.

    1978-04-01

    This test was conducted to complement Test VIII which was a sequential test of cables, cable splices, and insulation samples. In this test, the generic LOCA environments (radiation, temperature, pressure, chemical spray) were simulated and simultaneously applied to the test items. There were no failures of any assemblies and all were able to function at rated current and voltage throughout the entire test. An additional parameter, dissipation factor, was monitored in this test and when used in conjunction with capacitance, provided a better indication of insulation degradation.

  5. Calculations of multidimensional test reactor neutron diffusion problems with KORAT 3D code and its combination with two-velocity two-temperature thermohydraulic RATEG code

    Energy Technology Data Exchange (ETDEWEB)

    Voronova, O.A.; Grebennikov, A.N.; Zvenigorodskaya, O.A. [Russian Federal Nuclear Center, Arzamas (Russian Federation)] [and others

    1996-09-01

    The paper briefly presents the method for numerical computation of 3-D group neutron diffusion equation implemented in KORAT 3D code within SATURN complex and oriented to 3-D stationary and nonstationary calculations in nuclear reactor physics including those on up-to-date distributed-memory multiprocessors. The computational results are given obtained with the above mentioned code for multidimensional test reactor problems as well as the results of numerical efficiency studies for the code parallelization on multiprocessors Cray T3D and SP-2. Then the setup is given for 3-D dynamic problem simulating a hypothetical mode of WPBER reactor. The computational results are presented for the above problem which includes the neutron-nuclear processes, thermohydraulic processes and the feedback. The computations used two program complexes: KORAT 3D + multichannel two-velocity two-temperature thermohydraulics code RATEG and READY complex. (author)

  6. Application of thermohydraulic dispatcher in low temperature district heating systems for decreasing heat carrier transportation energy cost and increasing reliability of heat supply

    Science.gov (United States)

    Yavorovsky, Y. V.; Romanov, D. O.; Sennikov, V. V.; Sultanguzin, I. A.; Malenkov, A. S.; Zhigulina, E. V.; Lulaev, A. V.

    2017-11-01

    Low pressure district heating systems have low breakdown rate and allow decreasing heat carrier transportation energy cost by means of avoiding throttling of available water head. One of the basic elements of such systems is thermohydraulic dispatcher (THD) which separates primary circuit and secondary circuit (or circuits) that allows avoiding mutual hydraulic influence of circuits on each other and reducing water heads of network pumps. Analysis of perspective ways of using thermohydraulic dispatcher (THD) in low temperature district heating systems is made in this paper. Principal scheme and mathematical model of low pressure and temperature district heating system based on CHP generation with THD are considered. The main advantages of such systems are pointed out.

  7. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R.

    1976-07-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  8. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Directory of Open Access Journals (Sweden)

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  9. THE PREDICTION OF pH BY GIBBS FREE ENERGY MINIMIZATION IN THE SUMP SOLUTION UNDER LOCA CONDITION OF PWR

    Directory of Open Access Journals (Sweden)

    HYOUNGJU YOON

    2013-02-01

    Full Text Available It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3&4 and UCN 1&2. As results, pH of the sump solution for the SKN 3&4 was between 7.02 and 7.45, and for the UCN 1&2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl, HNO3, and Cs are very low.

  10. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    {sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.

  11. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  12. Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System.

    Science.gov (United States)

    Nemś, Magdalena; Nemś, Artur; Kasperski, Jacek; Pomorski, Michał

    2017-08-12

    This article presents the results of a study into a packed bed filled with ceramic bricks. The designed storage installation is supposed to become part of a heating system installed in a single-family house and eventually to be integrated with a concentrated solar collector adapted to climate conditions in Poland. The system's working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater. Designing a packed bed of sufficient parameters first required a mathematical model to be constructed and heat exchange to be analyzed, since heat accumulation is a complex process influenced by a number of material properties. The cases discussed in the literature are based on differing assumptions and different formulas are used in calculations. This article offers a comparison of various mathematical models and of system operating parameters obtained from these models. The primary focus is on the Nusselt number. Furthermore, in the article, the thermo-hydraulic efficiency of the investigated packed bed is presented. This part is based on a relationship used in solar air collectors with internal storage.

  13. Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

    Energy Technology Data Exchange (ETDEWEB)

    Arndt, S.A.

    1997-07-01

    The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.

  14. Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs; Comparacion de la fenomenologia de las secuencias de SBO con y sin LOCA de sellos en reactores PWR Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

    2013-07-01

    SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

  15. Analysis Thermo-hydraulic of trajectories related to procedures for operation of Emergency (POE). Application to the loss of a train of the DTH; Analisis termohidraulico de trayectorias vinculadas a Procedimientos de Operacion de emergencia (POE). Aplicacion a la perdida de un tren de RHR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Saez, F.; Martorell Alsina, S.; Carlos Alberola, A.; Villanueva Lopez, J. F.; Martorell Aygues, P.

    2012-07-01

    This work explores different possible sequences at the loss of a train of the DTH when the plant is lowering power. The study of the different possible trajectories has been done through the collapse tool and study thermo-hydraulic each of these paths is done by the code TRACE Thermo-hydraulic.

  16. Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code

    Science.gov (United States)

    Lobanov, P. D.; Usov, E. V.; Butov, A. A.; Pribaturin, N. A.; Mosunova, N. A.; Strizhov, V. F.; Chukhno, V. I.; Kutlimetov, A. E.

    2017-10-01

    Experiments with impulse gas injection into model coolants, such as water or the Rose alloy, performed at the Novosibirsk Branch of the Nuclear Safety Institute, Russian Academy of Sciences, are described. The test facility and the experimental conditions are presented in details. The dependence of coolant pressure on the injected gas flow and the time of injection was determined. The purpose of these experiments was to verify the physical models of thermohydraulic codes for calculation of the processes that could occur during the rupture of tubes of a steam generator with heavy liquid metal coolant or during fuel rod failure in water-cooled reactors. The experimental results were used for verification of the HYDRA-IBRAE/LM system thermohydraulic code developed at the Nuclear Safety Institute, Russian Academy of Sciences. The models of gas bubble transportation in a vertical channel that are used in the code are described in detail. A two-phase flow pattern diagram and correlations for prediction of friction of bubbles and slugs as they float up in a vertical channel and of two-phase flow friction factor are presented. Based on the results of simulation of these experiments using the HYDRA-IBRAE/LM code, the arithmetic mean error in predicted pressures was calculated, and the predictions were analyzed considering the uncertainty in the input data, geometry of the test facility, and the error of the empirical correlation. The analysis revealed major factors having a considerable effect on the predictions. The recommendations are given on updating of the experimental results and improvement of the models used in the thermohydraulic code.

  17. Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

    Energy Technology Data Exchange (ETDEWEB)

    Coragem, Helio Boemer de Oliveira

    1980-07-01

    A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

  18. Development and application of an innovative tool to automate the process of results extraction from the thermo-hydraulic simulator Olga

    Directory of Open Access Journals (Sweden)

    Francesco Carducci

    2015-06-01

    Full Text Available This paper presents the development and application of an innovative code to extract in an automated way data from the thermo-hydraulic simulator Olga. The results show that the tool can significantly reduce the time needed for the data extraction procedure and increase the reliability of results due to the fact that there is no more the need of the human operator. Moreover, during the data extraction phase, the Olga code is available for running different simulations allowing to optimize the use of this resource.

  19. Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW{sub th} power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H{sub 2} corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the

  20. Phosphorylation of the 12 S globulin cruciferin in wild-type and abi1-1 mutant Arabidopsis thaliana (thale cress) seeds

    Science.gov (United States)

    Wan, Lianglu; Ross, Andrew R. S.; Yang, Jingyi; Hegedus, Dwayne D.; Kermode, Allison R.

    2007-01-01

    Cruciferin (a 12 S globulin) is the most abundant storage protein in the seeds of Arabidopsis thaliana (thale cress) and other crucifers, sharing structural similarity with the cupin superfamily of proteins. Cruciferin is synthesized as a precursor in the rough endoplasmic reticulum. Subunit assembly is accompanied by structural rearrangements involving proteolysis and disulfide-bond formation prior to deposition in protein storage vacuoles. The A. thaliana cv. Columbia genome contains four cruciferin loci, two of which, on the basis of cDNA analysis, give rise to three alternatively spliced variants. Using MS, we confirmed the presence of four variants encoded by genes At4g28520.1, At5g44120.3, At1g03880.1 and At1g3890.1 in A. thaliana seeds. Two-dimensional gel electrophoresis, along with immunological detection using anti-cruciferin antiserum and antibodies against phosphorylated amino acid residues, revealed that cruciferin was the major phosphorylated protein in Arabidopsis seeds and that polymorphism far exceeded that predicted on the basis of known isoforms. The latter may be attributed, at least in part, to phosphorylation site heterogeneity. A total of 20 phosphorylation sites, comprising nine serine, eight threonine and three tyrosine residues, were identified by MS. Most of these are located on the IE (interchain disulfide-containing) face of the globulin trimer, which is involved in hexamer formation. The implications of these findings for cruciferin processing, assembly and mobilization are discussed. In addition, the protein phosphatase 2C-impaired mutant, abi1-1, was found to exhibit increased levels of cruciferin phosphorylation, suggesting either that cruciferin may be an in vivo target for this enzyme or that abi1-1 regulates the protein kinase/phosphatase system required for cruciferin phosphorylation. PMID:17313365

  1. Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments

    Energy Technology Data Exchange (ETDEWEB)

    Berta, V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-05-01

    Data from the Loss-of-Fluid Test (LOFT) Program help quantify the margin of safety inherent in pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). As early as 1979, questions arose concerning the accuracy of LOFT fuel rod cladding temperature data during several large-break LOCA experiments. This report analyzes how well externally-mounted fuel rod cladding thermocouples in LOFT accurately reflected actual cladding surface temperature during large-break LOCA experiments. In particular, the validity of the apparent core-wide fuel rod cladding quench exhibited during blowdown in LOFT Experiments L2-2 and L2-3 is studied. Also addressed is the question of whether the externally-mounted thermocouples might have influenced cladding temperature. The analysis makes use of data and information from several sources, including later, similar LOFT Experiments in which fuel centerline temperature measurements were made, experiments in other facilities, and results from a detailed FRAP-T6 model of the LOFT fuel rod. The analysis shows that there can be a significant difference (referred to as bias) between the surface-mounted thermocouple reading and the actual cladding temperature, and that the magnitude of this bias depends on the rate of heat transfer between the fuel rod cladding and coolant. The results of the analysis demonstrate clearly that a core-wide cladding quench did occur in Experiments L2-2 and L2-3. Further, it is shown that, in terms of peak cladding temperature recording during LOFT large-break LOCA experiments, the mean bias is 11.4 {plus_minus} 16.2K (20.5 {plus_minus} 29.2{degrees} F). The best-estimate value of peak cladding temperature for LOFT LP-02-6 is 1,104.8 K. The best-estimate peak cladding temperature for LOFT LP-LB-1 is 1284.0 K.

  2. Experimental investigation of effect of flow attack angle on thermohydraulic performance of air flow in a rectangular channel with discrete V-pattern baffle on the heated plate

    Directory of Open Access Journals (Sweden)

    Raj Kumar

    2016-05-01

    Full Text Available In this work, the effect of angle of attack ( α a of the discrete V-pattern baffle on thermohydraulic performance of rectangular channel has been studied experimentally. The baffle wall was constantly heated and the other three walls of the channel were kept insulated. The experimentations were conducted to collect the data on Nusselt number ( N u b and friction factor ( f b by varying the Reynolds number (Re = 3000–21,000 and angle of attack ( α a from 30° to 70°, for the kept values of relative baffle height ( H b / H = 0 . 50 , relative pitch ratio ( P b / H = 1 . 0 , relative discrete width ( g w / H b = 1 . 5 and relative discrete distance ( D d / L v = 0 . 67 . As compared to the smooth wall, the V-pattern baffle roughened channel enhances the Nusselt number ( N u b and friction factor ( f b by 4.2 and 5.9 times, respectively. The present discrete V-pattern baffle shapes with angle of attack ( α a of 60° equivalent to flow Reynolds number of 3000 yields the greatest thermohydraulic performance. Discrete V-pattern baffle has improved thermal performance as compared to other baffle shapes’ rectangular channel.

  3. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 2. Numerical investigations on proprieties of R/V upper plenum separation

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu; Yamaguchi, Akira [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Murakami, Satoshi [Customer System Co. Ltd., Tokai, Ibaraki (Japan)

    2002-03-01

    A large-scale sodium-cooled fast breeder reactor investigated in the feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to contribute to a planning of water scaled-model experiments for the separated upper plena of the reactor vessel. >From the analysis, the following results were obtained. (1) It can be considered that there is little effect of volumetric flow rate between the upper plenum and the free surface plenum on hydrodynamic characteristics in the whole upper plenum. Because the affected area is limited in the neighborhood of gaps of hot/cold leg pipes. (2) Change in flow velocity components is limited around the outer wall of hot leg pipes if the gap width of the dipped plate and penetrating components changes. From the above results, it was concluded that the assumption seems to be valid that the free surface plenum hydrodynamic characteristics is independent of the upper plenum flows. However it is necessary to consider jet effects due to the hot leg intake flows affecting vortex concentrations in the upper plenum. (author)

  4. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  5. Comparative experiments to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Li, YuQuan; Hao, Botao; Zhong, Jia; Wan Nam [State Nuclear Power Technology R and D Center, South Park, Beijing Future Science and Technology City, Beijing (China)

    2017-02-15

    The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME)—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative

  6. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

    Directory of Open Access Journals (Sweden)

    Li Yuquan

    2017-02-01

    Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

  7. An attempt for a unified description of mechanical testing on Zircaloy-4 cladding subjected to simulated LOCA transient

    Directory of Open Access Journals (Sweden)

    Desquines Jean

    2016-01-01

    Full Text Available During a Loss Of Coolant Accident (LOCA, an important safety requirement is that the reflooding of the core by the emergency core cooling system should not lead to a complete rupture of the fuel rods. Several types of mechanical tests are usually performed in the industry to determine the degree of cladding embrittlement, such as ring compression tests or four-point bending of rodlets. Many other tests can be found in the open literature. However, there is presently no real intrinsic understanding of the failure conditions in these tests which would allow translation of the results from one kind of mechanical testing to another. The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing. To achieve this objective, the failure mechanisms of pre-oxidized and pre-hydrided cladding samples are characterized by comparing the behavior of two different mechanical tests: Axial Tensile (AT test and “C”-shaped Ring Compression Test (CCT. The failure of samples in both cases can be described by usual linear elastic fracture mechanics theory. Using interrupted mechanical tests, metallographic examinations have evidenced that a set of parallel cracks are nucleated at the inner and outer surface of the samples just before failure, crossing both the oxide layer and the oxygen rich alpha layer. The stress intensity factors for multiple crack geometry are determined for both AT and CCT samples using finite element calculations. After each mechanical test performed on high temperature steam oxidized samples, metallography is then used to individually determine the crack depth and crack spacing. Using these two important parameters and considering the applied load at fracture, the stress intensity factor at failure is derived for each tested

  8. Geochemical and Hydrologic Controls of Copper-Rich Surface Waters in the Yerba Loca-Mapocho System

    Science.gov (United States)

    Pasten, P.; Montecinos, M.; Coquery, M.; Pizarro, G. E.; Abarca, M. I.; Arce, G. J.

    2015-12-01

    Andean watersheds in Northern and Central Chile are naturally enriched with metals, many of them associated to sulfide mineralizations related to copper mining districts. The natural and anthropogenic influx of toxic metals into drinking water sources pose a sustainability challenge for cities that need to provide safe water with the smallest footprint. This work presents our study of the transformations of copper in the Yerba Loca-Mapocho system. Our sampling campaign started from the headwaters at La Paloma Glacier and continues to the inlet of the San Enrique drinking water treatment plant, a system feeding municipalities in the Eastern area of Santiago, Chile. Depending on the season, total copper concentrations go as high as 22 mg/L for the upper sections, which become diluted to total reflection X ray fluorescence) and XRD (X-ray diffraction). Major elements detected in the precipitates were Al (200 g/kg), S (60 g/kg), and Cu (6 g/kg). Likely solid phases include hydrous amorphous phases of aluminum hydroxides and sulfates, and copper hydroxides/carbonates. Efforts are undergoing to find the optimal mixing ratios between the acidic stream and more alkaline streams to maximize attenuation of dissolved copper. The results of this research could be used for enhancing in-stream natural attenuation of copper and reducing treatment needs at the drinking water facility. Acknowledgements to Fondecyt 1130936 and Conicyt Fondap 15110020

  9. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  10. A Review of Dangerous Dust in Fusion Reactors: from Its Creation to Its Resuspension in Case of LOCA and LOVA

    Directory of Open Access Journals (Sweden)

    Andrea Malizia

    2016-07-01

    Full Text Available The choice of materials for the future nuclear fusion reactors is a crucial issue. In the fusion reactors, the combination of very high temperatures, high radiation levels, intense production of transmuting elements and high thermomechanical loads requires very high-performance materials. Erosion of PFCs (Plasma Facing Components determines their lifetime and generates a source of impurities (i.e., in-vessel tritium and dust inventories, which cool down and dilute the plasma. The resuspension of dust could be a consequences of LOss of Coolant Accidents (LOCA and LOss of Vacuum Accidents (LOVA and it can be dangerous because of dust radioactivity, toxicity, and capable of causing an explosion. These characteristics can jeopardize the plant safety and pose a serious threat to the operators. The purpose of this work is to determine the experimental and numerical steeps to develop a numerical model to predict the dust resuspension consequences in case of accidents through a comparison between the experimental results taken from campaigns carried out with STARDUST-U and the numerical simulation developed with CFD codes. The authors in this work will analyze the candidate materials for the future nuclear plants and the consequences of the resuspension of its dust in case of accidents through the experience with STARDUST-U.

  11. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu; Yamaguchi, Akira [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Murakami, Satoshi [Customer System Co.Ltd. (Japan)

    2002-02-01

    A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

  12. Thermohydraulics of a horizontal diphasic flow of superfluid helium; Thermo-hydraulique d'un ecoulement horizontal d'helium superfluide diphasique

    Energy Technology Data Exchange (ETDEWEB)

    Perraud, S

    2007-12-15

    This study aims at characterizing helium two phase flows, and to identify the dependence of their characteristics on various thermo-hydraulic parameters: vapour velocity, liquid height, vapour density, specificities of superfluidity. Both the engineer and the physicist's points of view are taken into consideration: the first one in terms of optimization of a particular cooling scheme based on a two-phase flow, and these second one in terms of more fundamental atomization-related questions. It has been shown that for velocities around 3 to 4 m/s, the liquid phase that was initially stratified undergoes an atomization through the presence of a drop haze carried by the vapor phase.This happens for superfluid helium as well as for normal helium without main differences on atomization.

  13. The InterFrost benchmark of Thermo-Hydraulic codes for cold regions hydrology - first inter-comparison results

    Science.gov (United States)

    Grenier, Christophe; Roux, Nicolas; Anbergen, Hauke; Collier, Nathaniel; Costard, Francois; Ferrry, Michel; Frampton, Andrew; Frederick, Jennifer; Holmen, Johan; Jost, Anne; Kokh, Samuel; Kurylyk, Barret; McKenzie, Jeffrey; Molson, John; Orgogozo, Laurent; Rivière, Agnès; Rühaak, Wolfram; Selroos, Jan-Olof; Therrien, René; Vidstrand, Patrik

    2015-04-01

    The impacts of climate change in boreal regions has received considerable attention recently due to the warming trends that have been experienced in recent decades and are expected to intensify in the future. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) that interact with the surrounding permafrost. For example, the thermal state of the surrounding soil influences the energy and water budget of the surface water bodies. Also, these water bodies generate taliks (unfrozen zones below) that disturb the thermal regimes of permafrost and may play a key role in the context of climate change. Recent field studies and modeling exercises indicate that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is required to understand and model the past and future evolution of landscapes, rivers, lakes and associated groundwater systems in a changing climate. However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, and the lack of study can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. Numerical approaches can only be validated against analytical solutions for a purely thermic 1D equation with phase change (e.g. Neumann, Lunardini). When it comes to the coupled TH system (coupling two highly non-linear equations), the only possible approach is to compare the results from different codes to provided test cases and/or to have controlled experiments for validation. Such inter-code comparisons can propel discussions to try to improve code performances. A benchmark exercise was initialized in 2014 with a kick-off meeting in Paris in November. Participants from USA, Canada, Germany, Sweden and France convened, representing altogether 13 simulation codes. The benchmark exercises consist of several test cases inspired by existing literature (e.g. McKenzie et al., 2007) as well as new ones. They

  14. Modifications in Compacted MX-80 Bentonite Due to Thermo-Hydraulic Treatment; Modificaciones en la Bentonita MX-80 Compactada Sometida a Tratamiento Termo-Hidraulico

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Espina, R.; Villar, M. V.

    2013-09-01

    The thermo-hydraulic tests reproduce the thermal and hydraulic conditions to which bentonite is subjected in the engineered barrier of a deep geological repository of radioactive waste. The results of thermo-hydraulic test TBT1500, which was running for approximately 1500 days, are presented. This is a continuation to the Technical Report Ciemat 1199, which presented results of test TBT500, performed under similar conditions but with duration of 500 days. In both tests the MX-80 bentonite was used with initial density and water content similar to those of the large-scale test TBT. The bentonite column was heated at the bottom at 140 degree centigrade and hydrated on top with deionized water. At the end of the test a sharp water content gradient was observed along the column, as well as an inverse dry density gradient. Hydration modified also the bentonite microstructure. Besides, an overall decrease of the smectite content with respect to the initial value took place, especially in the most hydrated areas where the percentage of interest ratified illite increased and in the longer test. On the other hand, the content of cristobalite, feldspars and calcite increased. Smectite dissolution processes (probably colloidal) occurred, particularly in the more hydrated areas and in the longer test. Due to the dissolution of low-solubility species and to the loss of exchangeable positions in the smectite, the content of soluble salts in the pore water increased with respect to the original one, especially in the longer test. The solubilized ions were transported; sodium, calcium, magnesium and sulphate having a similar mobility, which was in turn lower than that of potassium and chloride. The cationic exchange complex was also modified. (Author)

  15. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 3. Numerical investigation for thermal stratification phenomena in the upper plenum

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu; Yamaguchi, Akira [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Wakita, Junichi [Customer System Co. Ltd. (Japan)

    2002-06-01

    A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermal stratification characteristics in the upper plenum, and to investigate trade-off relations between gas entrainment and thermal stratification phenomena on in-vessel structures for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) Dummy plug insertion to a slit of the upper core structure is one of the effective measures to stabilize the in-vessel flow patterns and to mitigate in-vessel thermal shocks. (2) Though flow guide device such as a baffle ring attached to reactor vessel wall is an effective measure to eliminate impinging jet to dipped plate, rising characteristics of the thermal stratification interface are affected by the baffle ring devise. (3) Thermal stratification characteristics are not influenced very much by the installation of a partial inner barrel to the dipped plate, which is an effective measure to reduce the horizontal flow velocity components at free surface. (4) Labyrinth structures to the gap between the reactor vessel wall and the outer dipped plate have direct effects upon in-vessel thermal shock characteristics including thermal stratification phenomena due to the closing of flow path between the upper plenum and the free surface plenum. (author)

  16. Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of exceedance of damage by integrated Safety Analysis Methodology; Arboles de sucesos dinamicos aplicados a secuencias Full Spectrum LOCA. Calculo de la frequencia de excedencia del dano mediante la metodologia Analisis Integrados de Seguridad (ISA)

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

    2013-09-01

    The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Exceedance Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

  17. A procedure for addressing the fuel rod failures during LB-LOCA transient in Atucha-2 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina; Del Nevo, Alessandro; Parisi, Carlo; D' Auria, Francesco [University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Diotisalvi 2, 56122 Pisa (Italy); Mazzantini, Oscar [Nucleoelectrica Argentina S.A., UGCNAII, 2806 Lima (Argentina)

    2009-06-15

    the equilibrium burnup core is reached, the refueling is performed is order to move each FA just three times during its lifetime. Thus, a fresh FA is introduced into the core in a position and kept there until it reaches a certain burnup (transition burnup). Then it is moved to its final location, where it is burnt up to the discharge burnup. A peculiarity of the Atucha-2 design, common to other heavy water moderated reactors, is the positive void reactivity coefficient. This implies, in the case of a LB-LOCA event, the occurrence of a power peak at the very beginning of the transient due to the large void formation in the core channels. Thus the LOCA event is also a RIA event. Connection to Licensing: The Atucha-2 Construction License was issued in July 1981, upon a previously submitted Preliminary Safety Analysis Report (PSAR). Construction works were suspended in 1986 and resumed in 2005, with the objective to bring the plant into operation by the year 2010. To this aim, a twofold licensing analysis strategy is foreseen: the original safety design philosophy is preserved, and recent advances in nuclear safety technology are incorporated, as far as possible. Derived from the connected probabilistic approach, the double ended guillotine break (DEGB) is considered as a beyond design basis scenario (BDBA). Nevertheless, the demonstration of the design capability to withstand this event has an important role in the evaluation of the plant safety. A proposal for accident analysis has been developed by GRNSPG/UNIPI, starting from the original Siemens methodology and further enhanced by the inclusion of the use of modern best estimate computer codes and methods. The methodology includes also the evaluation of uncertainty in the calculated results, thus resulting in a Best Estimate plus Uncertainties approach (BEPU). Procedure addressing the fuel rod failures Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability

  18. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde; Simulacion de escenarios de accidente severo tipo perdida de refrigerante (Loca) pequeno, con el codigo MELCOR version 2.1 para la central nucleo-electrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft{sup 2} (4.6 cm{sup 2}), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  19. Analysis of the contribution and efficiency of the Santuario de la Naturaleza Yerba Loca, 33º S in protecting the regional vascular plant flora (Metropolitan and Fifth regions of Chile) Análisis de la contribución y eficiencia del Santuario de la Naturaleza Yerba Loca, 33º S, en la protección de la flora vascular regional (regiones Metropolitana y Quinta de Chile)

    OpenAIRE

    MARY T. K ARROYO; CLODOMIRO MARTICORENA; OSCAR MATTHEI; MÉLICA MUÑOZ; PATRICIO PLISCOFF

    2002-01-01

    Santuario de la Naturaleza Yerba Loca (SN Yerba Loca), Metropolitan Region (MR), 33º S, Chile is analyzed for its conservation value and efficiency in protecting native vascular plants in a regional context. The reserve's flora of 500 species and subtaxa was evaluated for species richness, endemism, range size and marginally distributed taxa, using species-area analysis, and tendencies in the floras of the MR (1.434 species and subtaxa) and MR-Fifth regions (1,841 species and subtaxa) to set ...

  20. Thermal-Hydraulic Assessment of W7-X Plasma Vessel Venting System in Case of 40 mm In-Vessel LOCA

    Directory of Open Access Journals (Sweden)

    E. Urbonavičius

    2015-01-01

    Full Text Available This paper presents assessment of the capacity of W7-X venting system in response to in-vessel LOCA, rupture of 40 mm diameter pipe during operation mode “baking.” The integral analysis of the coolant release from the cooling system, pressurisation of PV, and response of the venting system is performed using RELAP5 code. The same coolant release rate was introduced to the COCOSYS code, which is a lumped-parameter code developed for analysis of processes in containment of the light water reactors and the detailed analysis of the plasma vessel and the venting system is performed. Different options of coolant release modeling available in COCOSYS are compared to define the base case model, which is further used for assessment of the other parameters, that is, the failure of one burst disk, the temperature in the environment, and the pressure losses in the piping of venting system. The performed analysis identified the best option for coolant release modeling and showed that the capacity of the W7-X venting system is enough to prevent overpressure of the plasma vessel in the case of in-vessel LOCA.

  1. SB-LOCA beyond the design basis in a PWR experimental verification of am procedures in the PKL test facility

    Energy Technology Data Exchange (ETDEWEB)

    Mull, T.; Schoen, B.; Umminger, K.; Wegner, R. [Framatome ANP GmbH, Erlangen (Germany)

    2001-07-01

    The integral test facility PKL at the Technical Center of Framatome ANP (formerly Siemens/KWU) in Erlangen, Germany, simulates a 1300 MWe western type PWR. It is scaled by 1:145 in power and volume at original elevations. It features the entire primary side including four symmetrically arranged coolant loops and auxiliary and safety systems as well as the major part of the secondary side. The test series PKL III D, which was finished at the end of 1999, aimed at the exploration of safety margins and at the efficiency and optimization of operator initiated accident management (AM) procedures. Among others, several tests with small primary breaks combined with additional system failures were performed. This presentation describes test D3.1. The scenario under investigation was a small primary break (24 cm{sup 2} ) with simultaneous failure of the high pressure safety injection (HPSI), a beyond-design-basis scenario. For the German 1300 MWe PWRs, under such additional failure conditions, SB-LOCAs with leak sizes below 25 cm{sup 2} account for 18 % of the integral core damage frequency (CDF). This integral CDF can be estimated to be 3.1*10{sup -6} per year if no credit is taken from AM procedures. The break location in the test under consideration was in the cold leg between reactor coolant pump (RCP) and reactor pressure vessel (RPV). The assumed aggravating circumstances were HPSI failure and unavailability of 2 steam generators (SGs) as well as 3 out of 4 main steam relief and control valves (MS-RCV). The extra borating system was switched to injection mode at low pressurizer level but, by itself, would have been unable to maintain enough coolant to avoid core being uncovered before the pressure reached the setpoint of the accumulators (ACCs). The accident was managed by additional utilization of the chemical- and volume control system (CVCS) to inject water to partly neutralize the leak rate. The plant could be cooled down by 2 SGs using only one MS-RCV. The

  2. Analysis of the contribution and efficiency of the Santuario de la Naturaleza Yerba Loca, 33º S in protecting the regional vascular plant flora (Metropolitan and Fifth regions of Chile Análisis de la contribución y eficiencia del Santuario de la Naturaleza Yerba Loca, 33º S, en la protección de la flora vascular regional (regiones Metropolitana y Quinta de Chile

    Directory of Open Access Journals (Sweden)

    MARY T. K ARROYO

    2002-12-01

    Full Text Available Santuario de la Naturaleza Yerba Loca (SN Yerba Loca, Metropolitan Region (MR, 33º S, Chile is analyzed for its conservation value and efficiency in protecting native vascular plants in a regional context. The reserve's flora of 500 species and subtaxa was evaluated for species richness, endemism, range size and marginally distributed taxa, using species-area analysis, and tendencies in the floras of the MR (1.434 species and subtaxa and MR-Fifth regions (1,841 species and subtaxa to set the regional pattern. The reserve (0.7 % of MR land area and 0.3 % MR-Fifth land area contains 34 % of the MR and 27% of the MR-Fifth floras, and around 16-17 % of the mediterranean-climate area (regions IV-VIII flora of central Chile. Veech's Relative Richness Index (RRI revealed that SN Yerba Loca houses exaggerated richness in relation to its land area (28 % more species than expected from the regional model. However, endemism rates (35 % Continental Chile endemics, 22 % Mediterranean endemics, 3% MR-Vth endemics are statistically lower than in the MR (44 %, 29 %, 9 % and the MR-Vth (48 %, 31 %, 11 % floras, and SN Yerba Loca houses proportionately fewer MR endemics (2 % than the MR (6 %. Compared with the regional floras, the reserve contains statistically fewer marginally distributed species, and range size (median = five administrative regions is significantly larger. The reserve's outstanding species richness compensates for its low endemism rates bringing the absolute number of endemics to 92 % of the regional expectation. Corresponding values for marginally distributed species are 81 % (northern limits, 63% (southern limits and for median and shorter range taxa, 100 %. It is concluded that SN Yerba Loca is a highly efficient reserve from the point of view of vascular plant conservation, and represents an excellent conservation choice. SN Yerba Loca and MN El Morado (a second state protected area in the MR, conservatively, house 39 % of the native

  3. Las locas ilusiones

    OpenAIRE

    Mendívil,Julio

    2006-01-01

    Evidentemente vivimos descubriendo la pólvora, pues migración ha habido siempre. No sólo en los renglones míticos del éxodo bíblico ni en la triste ruta errabunda del pithecanthropus, también los desplazamientos de los visigodos por tierras sureñas o los asaltos de los hunos al este de Europa fueron migraciones. Puestos a ver las cosas, el mismo poblamiento del continente americano no fue sino el resultado de consecutivas hordas migratorias asiáticas llegadas al Nuevo Mundo por el estrecho de...

  4. LocaTag

    OpenAIRE

    Köbler, Felix;Koene, Philip;Goswami, Suparna;Leimeister, Jan Marco;Krcmar, Helmut

    2014-01-01

    In recent years instant messaging tools have been successfully introduced to support communication and collaboration processes in work environments. Research suggests that the use of instant messaging tools lead to an increased feeling of connectedness, social presence and awareness within a collaborative group. Consequently, the goal of our research is to improve communication and collaboration within a group through automated and real-time dissemination of presence information, using instan...

  5. Implementation of the optimization for the methodology of the neutronic calculation and thermo-hydraulic in IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo de; Conti, Thadeu das Neves; Fedorenko, Giuliana G.; Castro, Vinicius A.; Maio, Mireia F., E-mail: gstefani@ipen.b, E-mail: tnconti@ipen.b, E-mail: g.fedorenko@ipen.b, E-mail: vcastro@ipen.b, E-mail: mfmaio@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Santos, Thiago Augusto dos, E-mail: tsantos@ipen.b [Universidade de Sao Paulo (IFUSP), Sao Paulo, SP (Brazil). Inst. de Fisica

    2011-07-01

    This work objective was to create a manager program that would automate the programs and computer codes in use for neutronic calculation and thermo-hydraulic in IEA-R1 reactor thus making the process for calculation of safety parameters and for configuration change up to 98% faster than that used in the reactor today. This process was tested in combination with the reactor operators and is being implemented by the quality department. The main codes and programs involved in the calculations of configuration change are Leopard, Hammier-Technion, Twodb, Citation and Cobra. Calculations of delayed neutron and criticality coefficients given in the process of safety parameters calculation are given by the Hammer-Technion and Citation in a process that involves about eleven repetitions so that it meets all the necessary conditions (such different temperatures of the moderator and fuel). The results are entirely consistent with the expected and absolutely the same as those given by manual process. Thus the work shows its reliability as well the advantage of saving time, once a process that could take up to four hours was turned in one that takes around five minutes when done in a home computer. Much of this advantage is due to the fact that were created subprograms to treat the output of each program used and transform them into the input of the other programs, removing from it the intermediate essential data for this to occur, thus avoiding also a possible human error by handling the various data supplied. (author)

  6. Thermo-hydraulic modelling of the South East Gas Pipeline System - an integrated model; Modelagem termo-hidraulica do Sistema de Gasodutos do Sudeste : um modelo integrado

    Energy Technology Data Exchange (ETDEWEB)

    Vianna Neto, Armando M.; Santos, Arnaldo M.; Mercon, Eduardo G. [TRANSPETRO - PETROBRAS Transportes, Rio de Janeiro, RJ (Brazil)

    2003-07-01

    This paper presents the development of an integrated simulation model, for the numerical calculation of thermal-hydraulic behaviors in the Brazilian southeast onshore gas pipeline flow system, remotely operated by TRANSPETRO's Gas Pipeline Control Centre (CCG). In its final application, this model is supposed to provide simulated results at the closer range to reality, in order to improve gas pipeline simulation studies and evaluations for the system in question. Considering the fact that numerical thermo-hydraulic simulation becomes the CCG's most important tool to analyze the boundary conditions to adjust the mentioned gas flow system, this paper seeks and takes aim to the optimization of the following prime attributions of a gas pipeline control centre: verification of system behaviors, face to some unit maintenance stop or procedure, programmed or not, or to some new gas outlet or inlet connection to the system; daily operational compatibility analysis between programmed and realized gas volumes; gas technical expedition and delivery analysis. Finally, all this work was idealized and carried out within the one-phase flow domain (dry gas) (author)

  7. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  8. Development of an extended thermo-hydraulic simulation tool for fusion magnet design study - Application to the initial versions of JT-60SA TF coils layout

    Science.gov (United States)

    Nicollet, S.; Lacroix, B.; Zani, L.; Hertout, P.; Portafaix, Ch.; Villari, R.

    2010-01-01

    In the framework of the EU participation to JT-60SA project [1], a dedicated simulation tool named after Thermo-hydraulic EXtended Tool (TEXTO) was developed at CEA between 2006 and 2007 in order to address in a reliable way the calculation of the magnet conductor temperature increase and temperature margins in different operating conditions. The simulation process is based on three different codes, addressing each specific aspects (MCNP for the 3D nuclear heat calculation, TRAPS for magnetic field, ANSYS for 2D transverse thermal contribution of coil casing), which finally stand as input for the well established code GANDALF (with transient helium properties). Both steady-state operating and disruption transient regimes can be studied with this process and a first application is performed on the basis of the design and operating conditions available at this time on JT-60SA TF magnets, i.e. the first version of the different design stages. The complete analysis is shown together with the associated proposals for the TF conductor layout that could be derived from these studies.

  9. Feasibility study of the university of Utah TRIGA reactor power upgrade - part II: Thermohydraulics and heat transfer study in respect to cooling system requirements and design

    Directory of Open Access Journals (Sweden)

    Babitz Philip

    2013-01-01

    Full Text Available The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much more pronounced at higher power levels with vortex shedding also occurring. A departure from nucleate boiling analysis showed that while nucleate boiling begins near 210 kW it remains in this state and does not approach the critical heat flux at powers up to 500 kW. Based on these studies, two upgrades are proposed for extended operation and possibly higher reactor power level. Together with the findings from Part I studies, we conclude that increase of the reactor power is highly feasible yet dependable on its purpose and associated investments.

  10. Church ladies, good girls, and locas: stigma and the intersection of gender, ethnicity, mental illness, and sexuality in relation to HIV risk.

    Science.gov (United States)

    Collins, Pamela Y; von Unger, Hella; Armbrister, Adria

    2008-08-01

    Inner city women with severe mental illness may carry multiple stigmatized statuses. In some contexts these include having a mental illness, being a member of an ethnic minority group, being an immigrant, being poor, and being a woman who does not live up to gendered expectations. These potentially stigmatizing identities influence both the way women's sexuality is viewed and their risk for HIV infection. This qualitative study applies the concept of intersectionality to facilitate understanding of how these multiple identities intersect to influence women's sexuality and HIV risk. We report the firsthand accounts of 24 Latina women living with severe mental illness in New York City. In examining the interlocking domains of these women's sexual lives, we find that the women seek identities that define them in opposition to the stigmatizing label of "loca" (Spanish for crazy) and bestow respect and dignity. These identities have unfolded through the additional themes of "good girls" and "church ladies". Therefore, in spite of their association with the "loca", the women also identify with faith and religion ("church ladies") and uphold more traditional gender norms ("good girls") that are often undermined by the realities of life with a severe mental illness and the stigma attached to it. However, the participants fall short of their gender ideals and engage in sexual relationships that they experience as disempowering and unsatisfying. The effects of their multiple identities as poor Latina women living with severe mental illness in an urban ethnic minority community are not always additive, but the interlocking effects can facilitate increased HIV risks. Interventions should acknowledge women's multiple layers of vulnerability, both individual and structural, and stress women's empowerment in and beyond the sexual realm.

  11. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Energy Technology Data Exchange (ETDEWEB)

    Salesa, F.

    2014-02-01

    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  12. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  13. Geophysical investigation and monitoring of thermo-hydraulic conditions of closed talik and icing of the Kuuguluk River at Salluit, northern Quebec, Canada

    Science.gov (United States)

    Fortier, R.; Lemieux, J. M.; Molson, J. W. H.; Therrien, R.; Ouellet, M.

    2016-12-01

    The Inuit community of Salluit in northern Quebec, Canada, is located in the continuous permafrost zone characterized by a mean annual air temperature (MAAT) of -8.0 °C over the period from 1981 to 2010. In such cold environment, it is challenging to find a sustainable supply of water. A well drilled in fractured bedrock and located in a closed talik underneath the Kuuguluk River is used as a source of drinking water by the municipality of Salluit. To verify the lateral extent of the closed talik beneath the floodplain of Kuuguluk River, a geophysical investigation using ground penetrating radar (GPR) profiling and capacitively-coupled electrical resistivity tomography (ERT) was undertaken in spring 2011. Moreover, a mooring with water level and temperature dataloggers in the river was installed over the 2015-2016 period to assess the thermo-hydraulic conditions of the river bed. The icing which forms each year in the floodplain of Kuuguluk River was used in spring 2011 as a bridge to cross over the river and move along the geophysical equipment. Three thaw bulbs in the ice-rich permafrost of the floodplain were inferred from low resistivity anomalies in the model of electrical resistivity. The largest bulb is about 40 m wide and 14 m thick. According to the mooring results, the mean annual temperature of the river bed (MATRB) was 1.4 °C in 2015-2016 while the MAAT was -7.1 °C. This MATRB above 0 °C is due to the heat storage of running surface water in the river bed and the suprapermafrost water flow in the closed talik. River bed temperature below 0 °C and as low as -3 °C from October 10th 2015 to November 20th 2015 and from January 23rd to April 17th 2016 were recorded. The spring freshet occurred on June 24th2016. Outside these periods, the river bed temperature stayed remarkably stable at 0.05 °C in winter time. While the water level in the Kuuguluk River varies from 0.4 to 1.0 m in summer time following the precipitation events, the water pressure can

  14. 3D thermohydraulic modeling of the coupling of surface water bodies to the subsurface below the major urban center of Berlin

    Science.gov (United States)

    Frick, Maximilian; Scheck-Wenderoth, Magdalena; Cacace, Mauro; Schneider, Michael

    2017-04-01

    This study aims at a better understanding of the present-day thermal and hydraulic configuration below the major urban center of Berlin, capital city of Germany. The study area is located in the Northeast German Basin, showing an infill of several kilometers of sediments. Herein, the shallow sedimentary succession is made up of a sequence of alternating aquifers and aquitards, most importantly the local aquitard of the Rupelian clay. This geological unit represents a natural barrier between the deeper saline aquifers and the shallow fresh water aquifers from whom Berlin produces 100% of its drinking water. Additionally, the shallow thermal and hydraulic configuration has been anthropogenically overprinted which may also influence deeper domains to some extent. In this study we make use of 3D thermohydraulic models of the subsurface, focusing on the coupling of surface water bodies to the underground, based on newly available hydraulic data integrated into a 3D hydrogeological model. The results of the study show, that the coupling of surface water bodies and groundwater might lead to significant modifications of predicted subsurface temperatures and fluid flow field. These modifications are most prominent, where differences in hydraulic head between surface water bodies and the adjacent aquifers are highest. Consequently, the predicted surface to groundwater flow field differs most in these areas and it also results in differences in predicted temperatures as a consequence of advective heat transport. Quantitatively, the presence of major lakes may account for temperature differences up to 5°C, while considering rivers only accounts for modifications up to 1°C. Additionally, the models created in this study set up a basis for future thermohaline simulations as saline groundwater may represent a threat to drinking water supply. First results from the models run in this study already indicate, that uprising heated water from deeper domains may rise to shallow

  15. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (LIMA/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m{sup 3} of soda almost instantaneously was lost. This inventory contained 1.87x10{sup 10} Bq/m{sup 3} of tritium, 2.22x10{sup 7} Bq/m{sup 3} of cobalt,3.48x10{sup 8} Bq/m{sup 3} of cesium and 3.44x10{sup 10} Bq/m{sup 3} of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10{sup 6} Bq/m{sup 3}, 1,11x10{sup 4} Bq/m{sup 3} and 1,85x10{sup 3} Bq/m{sup 3}) after 22 hours, respectively for {sup 3}H, {sup 60}Co, {sup 131}I and {sup 137}Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  16. Unités discursives de base et leur périphérie gauche dans LOCAS-F, un corpus oral multigenres annoté

    Directory of Open Access Journals (Sweden)

    Degand Liesbeth

    2014-07-01

    Full Text Available Cette contribution vise à présenter le corpus LOCAS-F (Louvain Corpus of Annotated Speech - French, développé à Louvain-la-Neuve (Belgique depuis 2011 et annoté au niveau discursif. Au coeur de l’annotation se trouve l’unité discursive de base résultant de la corrélation entre unités prosodiques et unités syntaxiques (voir notamment Lacheret-Dujour et Victorri, 2002. Partant du postulat que la syntaxe et la prosodie fournissent chacune des signaux pour la délimitation d’unités dans le flux de parole, nous proposons que l’unité discursive soit définie par la coïncidence entre frontières syntaxique et prosodique. Le premier volet de cette contribution vise à présenter notre définition de ces unités, ainsi que notre méthode de segmentation, qui mêle une annotation syntaxique et une annotation prosodique, réalisées de manière totalement indépendante. Nous présentons ensuite l’ensemble du corpus LOCAS-F et ses différentes composantes, en nous attardant sur la distribution des unités discursives de base au sein des différents genres constituant notre corpus et justifions leur usage par rapport à la distribution des unités syntaxiques et des unités prosodiques seules. Nous poursuivons en effet l’hypothèse (en section 2 selon laquelle seule la combinaison de ces deux types d’unités permet de définir des unités pertinentes pour l’analyse du discours, remplissant une fonction cognitive dans la planification et l’interprétation de celui-ci. Enfin, nous présentons une étude exploitant le corpus annoté en nous concentrant sur les amorces des unités discursives. La périphérie gauche constitue le point d’ancrage du message et est le lieu, selon nous, où les locuteurs peuvent signaler la macro-structure de leur discours. Afin d’étudier les stratégies discursives à l’œuvre dans notre corpus, nous établissons dès lors une analyse distributionnelle des types formels observés et, sur un

  17. Lugar de encuentros de tópicos románicos : Doña Juana la Loca de Pradilla

    Directory of Open Access Journals (Sweden)

    José Enrique García Melero

    1999-01-01

    Full Text Available Estudio del cuadro titulado «Doña Juana la Loca» de Pradilla, que aquí es considerado como la culminación de la pintura de historia en España y el inicio de su decadencia. Se atribuye su éxito en su época a la concurrencia en él de una serie de características románticas tanto formales como de contenido: la representación de la locura de una Reina enamorada y celosa debido a la muerte de su esposo el Rey Felipe I, tema sacado de la Historia al poco tiempo de acabar la Edad Media y en el inicio del Imperio español. Pero, además, se destacan otros aspectos importantes del cuadro: su escenografía tan teatral y romántica, la representación del amanecer de un día invernal de cielo nublado, el fuego y el humo de una hoguera, el árbol seco, el viento, la figura enlutada de la Reina... Se considera al cuadro como un manierismo del género, como una importante consecuencia del excesivo intelectualismo al que había llegado la pintura de historia hacia 1875. Se establece una relación con el poema de García Lorca titulado «Elegía a Doña Juana la Loca».Learning of the Pradilla's painting titled «Mrs Juana the Lunatic», which is considered here as the culmination of the historical painting in Spain and the commence of its decline. The success of Pradilla's painting in his epoch is attributed to the conjunction in it of a set of characterístics related not only to its forms, but also to its contents: the representation of a Queen who is in love and jealous, and became insane due to the death of her husband, Felipe I. This theme is inspired in the History, little time after the end of the Middle Ages and at the beginning of the Spanish Empire. Besides, there are another aspects of the painting emphasised, such as its theatrical and romantic scenography: the representation of the dawn in a wintry and cloudy day, the fire and smoke caused by a bonfire, the dead tree, the wind, the Queen's in mourning shape... This painting is

  18. Transient thermohydraulic heat pipe modeling

    Science.gov (United States)

    Hall, Michael L.; Doster, Joseph M.

    Many space based reactor designs employ heat pipes as a means of conveying heat. In these designs, thermal radiation is the principle means for rejecting waste heat from the reactor system, making it desirable to operate at high temperatures. Lithium is generally the working fluid of choice as it undergoes a liquid-vapor transformation at the preferred operating temperature. The nature of remote startup, restart, and reaction to threats necessitates an accurate, detailed transient model of the heat pipe operation. A model is outlined of the vapor core region of the heat pipe which is part of a large model of the entire heat pipe thermal response. The vapor core is modeled using the area averaged Navier-Stokes equations in one dimension, which take into account the effects of mass, energy and momentum transfer. The core model is single phase (gaseous), but contains two components: lithium gas and a noncondensible vapor. The vapor core model consists of the continuity equations for the mixture and noncondensible, as well as mixture equations for internal energy and momentum.

  19. Simulación de operación de una central nuclear con el simulador gráfico interactivo: LOCA en rama fría en BOL, MOL y EOL

    OpenAIRE

    Penche Alonso, Marta

    2016-01-01

    La energía nuclear convencional se basa en reacciones de fisión producidas cuando los átomos de uranio absorben neutrones. Se tratan de reacciones autosostenidas en el combustible del reactor nuclear. Este trabajo se enmarca dentro del campo de la seguridad nuclear, estudiando, mediante simulaciones, el desarrollo del accidente base de diseño LOCA. En la seguridad de una central nuclear hay varios niveles: - Primer nivel: la prevención de accidentes. - Segundo nivel: el control de la...

  20. Taxonomy Icon Data: thale cress [Taxonomy Icon

    Lifescience Database Archive (English)

    Full Text Available .png Arabidopsis_thaliana_S.png Arabidopsis_thaliana_NS.png http://biosciencedbc.jp/taxonomy_icon/icon.cgi?i...=Arabidopsis+thaliana&t=L http://biosciencedbc.jp/taxonomy_icon/icon.cgi?i=Arabidopsis+thaliana&t=NL http://...biosciencedbc.jp/taxonomy_icon/icon.cgi?i=Arabidopsis+thaliana&t=S http://biosciencedbc.jp/taxonomy_icon/icon.cgi?i=Arabidopsis+thaliana&t=NS ...

  1. Thales of Miletus (624-546 BC)

    Science.gov (United States)

    Murdin, P.

    2000-11-01

    The first known Greek philosopher; also a scientist, mathematician and engineer, born in Miletus, Asia Minor (now Turkey), believed to have been the teacher of ANAXIMANDER. Apparently wrote a book on navigation, defining the constellation Ursa Minor and using it as a navigation aid. Credited with the prediction of an eclipse of the Sun in 585 BC, although it is not known how, since the Metonic cy...

  2. Multiplication: From Thales to Lie1

    Indian Academy of Sciences (India)

    geometric construction of addi- tion to that of multiplication in a presentation given to high school students on the occasion of the. Science Day (28-2-2008) in. BITS, Pilani. The unraveling of this magic is the genesis of this article. sides of a parallelogram have the same length. In effect, the shadow of a shadow transports the ...

  3. Importance of water quality on plant abundance and diversity in high-alpine meadows of the Yerba Loca Natural Sanctuary at the Andes of north-central Chile Importancia de la calidad del agua sobre la abundancia y diversidad vegetal en vegas altoandinas del Santuario Natural Yerba Loca en los Andes de Chile centro-norte

    Directory of Open Access Journals (Sweden)

    ROSANNA GINOCCHIO

    2008-12-01

    Full Text Available Porphyry Cu-Mo deposits have influenced surface water quality in high-Andes of north-central Chile since the Miocene. Water anomalies may reduce species abundance and diversity in alpine meadows as acidic and metal-rich waters are highly toxic to plants The study assessed the importance of surface water quality on plant abundance and diversity in high-alpine meadows at the Yerba Loca Natural Santuary (YLNS, central Chile (33°15' S, 70°18' W. Hydrochemical and plant prospecting were carried out on Piedra Carvajal, Chorrillos del Plomo and La Lata meadows the growing seasons of 2006 and 2007. Direct gradient analysis was performed through canonical correspondence analysis (CCA to look for relationships among water chemistry and plant factors. High variability in water chemistry was found inside and among meadows, particularly for pH, sulphate, electric conductivity, hardness, and total dissolved Cu, Zn, Cd, Pb and Fe. Data on species abundance and water chemical factors suggests that pH and total dissolved Cu are very important factor determining changes in plant abundance and diversity in study meadows. For instance, Festuca purpurascens, Colobanthus quitensis, and Arenaria rivularis are abundant in habitals with Cu-rich waters while Festuca magellanica, Patosia clandestina, Plantago barbata, Werneria pygmea, and Erigeron andícola are abundant in habitals with dilute waters.Los megadepósitos de pórfidos de Cu-Mo han influido sobre la calidad de las aguas superficiales en las zonas altoandinas del centro-norte de Chile desde el Mioceno. Estas alteraciones en la calidad de las aguas podrían afectar negativamente a la vegetación presente en las vegas altoandinas, ya que las aguas acidas y ricas en metales son altamente tóxicas para las plantas. En este estudio se evaluó el efecto de la calidad de las aguas en la abundancia y diversidad florística de las vegas altoandinas del Santuario de la Naturaleza Yerba Loca (SNYL, en Chile central (33

  4. Study of the spatial dependence of neutronic flow oscillations caused by fluctuations thermohydraulics at the entrance of the core of a reactor PWR; Estudio de la dependencia espacial de las oscilaciones de flujo neutronico causadas por flucturaciones termohidraulicas a la entrada del nucleo de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bermejo, J. A.; Lopez, A.; Ortego, A.

    2014-07-01

    It presents a theoretical study on spatial dependence of flow oscillations neutronic caused by thermal hydraulics fluctuations at the entrance of the core of a PWR reactor. To simulate, with SIMULATE code - 3K different fluctuations thermohydraulics at the entrance to the core and the spatial dependence of the oscillations and is analyzed neutronic flow obtained at locations of neutron detectors. the work It is part of the r and d program initiated in CNAT to investigate the phenomenon of the noise neutronic. (Author)

  5. Selva, plumas y desconche : Un análisis de las performances masculinas de la feminidad entre las locas del Tigre durante la década del ochenta.

    Directory of Open Access Journals (Sweden)

    Santiago Joaquin Insausti

    2010-08-01

    Full Text Available En este trabajo me propongo describir los modos en que las personas que ejercían géneros o sexualidades diversas construían sus subjetividades y sus cuerpos en Argentina en las décadas del setenta y el ochenta. Para esto, partiré del análisis de las fotos que estas personas se tomaban en el contexto de fiestas clandestinas organizadas en las islas del Tigre. Partiendo de la tesis constructivista de la ausencia de distancia entre género y expresión de género, indagare en las fotos la relación de las imágenes con el contexto selvático que las enmarca, la construcción contrapuesta de chongos y locas, dos de las principales identidades sexuales de la época, y en los usos del cuerpo utilizados por estos para producir feminidad y masculinidad. Finalmente, arriesgare dos hipótesis tendientes a explicar las causas de la masiva migración de estos sujetos desde la ciudad hacia el delta del Paraná.

  6. Selva, plumas y desconche : Un análisis de las performances masculinas de la feminidad entre las locas del Tigre durante la década del ochenta.

    Directory of Open Access Journals (Sweden)

    Santiago Joaquin Insausti

    2011-08-01

    Full Text Available En este trabajo me propongo describir los modos en que las personas que ejercían géneros o sexualidades diversas construían sus subjetividades y sus cuerpos en Argentina en las décadas del setenta y el ochenta. Para esto, partiré del análisis de las fotos que estas personas se tomaban en el contexto de fiestas clandestinas organizadas en las islas del Tigre. Partiendo de la tesis constructivista de la ausencia de distancia entre género y expresión de género, indagare en las fotos la relación de las imágenes con el contexto selvático que las enmarca, la construcción contrapuesta de chongos y locas, dos de las principales identidades sexuales de la época, y en los usos del cuerpo utilizados por estos para producir feminidad y masculinidad. Finalmente, arriesgare dos hipótesis tendientes a explicar las causas de la masiva migración de estos sujetos desde la ciudad hacia el delta del Paraná.

  7. Thales: His contribution to scientific knowledge | Asukwo | Sophia ...

    African Journals Online (AJOL)

    The cosmocentic period of Western Philosophy was a child of 'circumstance'. This was because of the general dissatisfaction from Homer and Hesiod's explanations of the universe. Their accounts were based on mythology, which could not attract the attention and the fancy of the people for long. The cosmoscentic period of

  8. A history of astronomy from Thales to Kepler

    CERN Document Server

    Dreyer, J L E

    1953-01-01

    A masterpiece of historical insight and scientific accuracy, this is the definitive work on Greek astronomy and the Copernican Revolution. Beginning with the ancient Egyptians, it ranges from the Pythagoreans and Plato to medieval European and Islamic cosmologies, concluding with detailed surveys of the works of Copernicus, Brahe, and Kepler.

  9. Church ladies, good girls, and locas

    Science.gov (United States)

    Collins, Pamela Y; von Unger, Hella; Armbrister, Adria

    2008-01-01

    Inner city women with severe mental illness may carry multiple stigmatized statuses. In some contexts these include having a mental illness, being a member of an ethnic minority group, being an immigrant, being poor, and being a woman who does not live up to gendered expectations. These potentially stigmatizing identities influence both the way women’s sexuality is viewed and their risk for HIV infection. This qualitative study applies the concept of intersectionality to facilitate understanding of how these multiple identities intersect to influence women’s sexuality and HIV risk. We report the firsthand accounts of 24 Latina women living with severe mental illness in New York City. In examining the interlocking domains of these women’s sexual lives, we find that the women seek identities that define them in opposition to the stigmatizing label of “loca” (Spanish for crazy) and bestow respect and dignity. These identities have unfolded through the additional themes of “good girls” and “church ladies”. Therefore, inspite of their association with the “loca”, the women also identify with faith and religion (“church ladies”) and uphold more traditional gender norms (“good girls”) that are often undermined by the realities of life with a severe mental illness and the stigma attached to it. However, the participants fall short of their gender ideals and engage in sexual relationships that they experience as disempowering and unsatisfying. The effects of their multiple identities as poor Latina women living with severe mental illness in an urban ethnic minority community are not always additive, but the interlocking effects can facilitate increased HIV risks. Interventions should acknowledge women’s multiple layers of vulnerability, both individual and structural, and stress women’s empowerment in and beyond the sexual realm. PMID:18423828

  10. Control of the neutronic and thermohydraulic conditions of power ramps in an irradiation loop for PWR fuel rod; Controle des conditions neutroniques et thermohydrauliques des rampes de puissance dans une boucle d`irradiation de combustibles de reacteur a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Moulin, D.J.F.

    1993-09-10

    In order to study the power transients effects on PWR fuel rod clad, ramp tests in a pressurized water loop, are carried out at OSIRIS reactor. The present thesis deals with the on-line control of the device, during power ramp and conditioning irradiation. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation of the Silver self-powered neutron detector was developed. With this method, the uncertainty of the ramp end-point is lower than 1%, thus it is very suited for monitoring both transient, as well as steady state conditions. Furthermore, a thermohydraulic model of the irradiation device is described: heat transfer equations, including gamma heating in materials, are solved to obtain temperatures and thermal fluxes of steady states. Results from the model and temperature measurements of the coolant are used together for fuel power determination, in real time. The clad external temperature profile is also calculated and displayed, to improve the irradiation monitoring. (author), 51 refs., 12 annexes, 66 figs.

  11. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  12. Neural networks for the analysis of margins of safety through code BE+U; Redes neuronales para el analisis de margenes de seguridad mediante codigos be+u

    Energy Technology Data Exchange (ETDEWEB)

    Villamizar, M.; Martorell, S.; Villanueva, J. F.; Carlos, S.; Sanchez, A.; Serradell, V.; Mendizabal, R.; Pelayo, F.; Sol, I.

    2011-07-01

    This paper presents the use of tools {sup S}oft Computing{sup ,} in particular the use of artificial neural networks and the method of decomposition of variance as sensitivity analysis, which allows understanding and modeling the relations between variables uncertain input inputs (defined by functions of distribution of Thermo-hydraulic model parameters) and output outputs variable presentation takes place on LOCA accident in a PWR as application.

  13. EL "MAL DE LAS VACAS LOCAS": UN TEMA DE BIOÉTICA EN LOS NUEVOS ESCENARIOS O "MAL DAS VACAS LOUCAS": UM TEMA DE BIOÉTICA NOS NOVOS CENÁRIOS THE "MAD COW DISEASE": A BIOETHICS’ ISSUE IN THE NEW SETTINGS

    Directory of Open Access Journals (Sweden)

    José Miguel Vera Lara

    2001-01-01

    Full Text Available La encefalopatía espongiforme bovina ("mal de las vacas locas", se produce en un momento y en un contexto que permiten resaltar los éxitos y fracasos de la biotecnología. En el marco de la globalización ha quedado de manifiesto la necesidad de establecer límites para la tecnociencia que en materia de logros aparecía como ilimitada. Se ha puesto en la balanza la rentabilidad de la industria cárnica versus la salud pública. El mal de las vacas locas constituye un traspié que no fue previsto, porque no era previsible en la perfectibilidad del sistema, lo que ha dejado al descubierto una fisura que, por ahora, es contextual. El contagio de las vacas a las personas es una campana de alarma, una advertencia que debe ser escuchada. La crisis esta recién en su etapa inicialA encefalopatia espongiforme bovina (mal da vaca louca ocorre em momento e contexto que permitem ressaltar êxitos e fracassos da biotecnologia. O marco da globalização tornou clara a necessidade de estabelecer limites para a tecnociência que por seus feitos mostrava-se ilimitada. Colocou-se na balança a rentabilidade da indústria da carne versus a saúde pública. O mal das vacas loucas constitui um deslize imprevisto, já que não era esperado considerando-se a perfeição do sistema e expos assim uma fissura que mostra-se contextual. O contágio de seres humanos é um alerta, uma advertência que deve ser considerada. A crise está apenas em sua etapa inicialBovine Spongiform Encephalopathy ("mad cow disease" appears in a moment and in a context that highlights the achievements and failures of biotechnology. The reality of globalization evidences the need to set limits to a technoscience viewed as unlimited because of its breakthroughs. Livestock industry profitability versus public health have been placed on the balance. Mad cow disease is an unforeseen stumble, since it was something unexpected in a system considered as perfect. This fact has disclosed a fissure

  14. RBMK thermohydraulic safety assessments using RELAP5/MOD3 codes

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.V.; Schmitt, B.E.

    1995-06-01

    The capability of the RELAP5/MOD3 code to validate various transients encountered in RBMK reactor postulated accidents has been assessed. The assessment results include a loss of coolant accident at the inlet of the core pressure tube, the blockage of a pressure tube, and the pressure response of the core cavity to in core pressure tube ruptures. These assessments show that the RELAP5/MOD3 code can predict major phenomena during postulated accidents in the RBMK reactors.

  15. A Thermo-Hydraulic Tool for Automatic Virtual Hazop Evaluation

    Directory of Open Access Journals (Sweden)

    Pugi L.

    2014-12-01

    Full Text Available Development of complex lubrication systems in the Oil&Gas industry has reached high levels of competitiveness in terms of requested performances and reliability. In particular, the use of HazOp (acronym of Hazard and Operability analysis represents a decisive factor to evaluate safety and reliability of plants. The HazOp analysis is a structured and systematic examination of a planned or existing operation in order to identify and evaluate problems that may represent risks to personnel or equipment. In particular, P&ID schemes (acronym of Piping and Instrument Diagram according to regulation in force ISO 14617 are used to evaluate the design of the plant in order to increase its safety and reliability in different operating conditions. The use of a simulation tool can drastically increase speed, efficiency and reliability of the design process. In this work, a tool, called TTH lib (acronym of Transient Thermal Hydraulic Library for the 1-D simulation of thermal hydraulic plants is presented. The proposed tool is applied to the analysis of safety relevant components of compressor and pumping units, such as lubrication circuits. Opposed to the known commercial products, TTH lib has been customized in order to ease simulation of complex interactions with digital logic components and plant controllers including their sensors and measurement systems. In particular, the proposed tool is optimized for fixed step execution and fast prototyping of Real Time code both for testing and production purposes. TTH lib can be used as a standard SimScape-Simulink library of components optimized and specifically designed in accordance with the P&ID definitions. Finally, an automatic code generation procedure has been developed, so TTH simulation models can be directly assembled from the P&ID schemes and technical documentation including detailed informations of sensor and measurement system.

  16. Thermo-hydraulic performance enhancement of solar air heater ...

    African Journals Online (AJOL)

    DR OKE

    sources of energy. Solar energy has a potential to fulfill the energy requirements of all human made systems provided technologies are developed to tap the potential of solar energy (Chamoli, 2013). Considerable efforts are being made to develop technologies to tap the great potential of solar energy. Air is generally used ...

  17. Combined Thermo-Hydraulic Analysis of a Cryogenic Jet

    CERN Document Server

    Chorowski, M

    1999-01-01

    A cryogenic jet is a phenomenon encountered in different fields like some technological processes and cryosurgery. It may also be a result of cryogenic equipment rupture or a cryogen discharge from the cryostats following resistive transition in superconducting magnets. Heat exchange between a cold jet and a warm steel element (e.g. a buffer tank wall or a transfer line vacuum vessel wall) may result in an excessive localisation of thermal strains and stresses. The objective of the analysis is to get a combined (analytical and experimental) one-dimensional model of a cryogenic jet that will enable estimation of heat transfer intensity between the jet and steel plate with a suitable accuracy for engineering applications. The jet diameter can only be determined experimentally. The mean velocity profile can be calculated from the fact that the total flux of momentum along the jet axis is conserved. The proposed model allows deriving the jet crown area with respect to the distance from the vent and the mean veloc...

  18. Numerical modeling of secondary side thermohydraulics of horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Melikhov, V.I.; Melikhov, O.I.; Nigmatulin, B.I. [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)

    1995-12-31

    A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.

  19. Thermo-hydraulic performance enhancement of solar air heater ...

    African Journals Online (AJOL)

    The present paper studies the thermal performance of solar air heater which is artificially roughened by providing multiple arcs with gap shaped roughness element. As thermal efficiency of smooth collector is quite low, hence there is a need to augment heat transfer from the absorbing surface. The experimentation has ...

  20. Thermo-hydraulic performance enhancement of solar air heater ...

    African Journals Online (AJOL)

    DR OKE

    The present paper studies the thermal performance of solar air heater which is artificially roughened by providing multiple arcs with gap shaped ... Keywords: Solar air heater; Nusselt number; thermal efficiency; multiple arcs with gap; roughened .... The glass wool was used as insulation inside wooden panel to reduce.

  1. An analytical thermohydraulic model for discretely fractured geothermal reservoirs

    Science.gov (United States)

    Fox, Don B.; Koch, Donald L.; Tester, Jefferson W.

    2016-09-01

    In discretely fractured reservoirs such as those found in Enhanced/Engineered Geothermal Systems (EGS), knowledge of the fracture network is important in understanding the thermal hydraulics, i.e., how the fluid flows and the resulting temporal evolution of the subsurface temperature. The purpose of this study was to develop an analytical model of the fluid flow and heat transport in a discretely fractured network that can be used for a wide range of modeling applications and serve as an alternative analysis tool to more computationally intensive numerical codes. Given the connectivity and structure of a fracture network, the flow in the system was solved using a linear system of algebraic equations for the pressure at the nodes of the network. With the flow determined, the temperature in the fracture was solved by coupling convective heat transport in the fracture with one-dimensional heat conduction perpendicular to the fracture, employing the Green's function derived solution for a single discrete fracture. The predicted temperatures along the fracture surfaces from the analytical solution were compared to numerical simulations using the TOUGH2 reservoir code. Through two case studies, we showed the capabilities of the analytical model and explored the effect of uncertainty in the fracture apertures and network structure on thermal performance. While both sources of uncertainty independently produce large variations in production temperature, uncertainty in the network structure, whenever present, had a predominant influence on thermal performance.

  2. Thermohydraulic safety issues for liquid metal cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Gerbeth, Gunter; Stefani, Frank [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany). Inst. of Fluid Dynamics; Eckert, Sven

    2016-05-15

    In this paper recent developments of various techniques for single-phase and two-phase flow measurements with relevance to liquid metal cooled systems will be presented. Further, the status of the DRESDYN platform for large-scale experiments with liquid sodium is sketched.

  3. Heaven and Earth in Ancient Greek Cosmology From Thales to Heraclides Ponticus

    CERN Document Server

    Couprie, Dirk L

    2011-01-01

    In Miletus, about 550 B.C., together with our world-picture cosmology was born. This book tells the story. In Part One the reader is introduced in the archaic world-picture of a flat earth with the cupola of the celestial vault onto which the celestial bodies are attached. One of the subjects treated in that context is the riddle of the tilted celestial axis. This part also contains an extensive chapter on archaic astronomical instruments. Part Two shows how Anaximander (610-547 B.C.) blew up this archaic world-picture and replaced it by a new one that is essentially still ours. He taught that the celestial bodies orbit at different distances and that the earth floats unsupported in space. This makes him the founding father of cosmology. Part Three discusses topics that completed the new picture described by Anaximander. Special attention is paid to the confrontation between Anaxagoras and Aristotle on the question whether the earth is flat or spherical, and on the battle between Aristotle and Heraclid...

  4. WHY THE WATER? THE VISION OF THE WORLD BY THALES OF MILETUS

    OpenAIRE

    KORCZAK, Andrzej

    2013-01-01

    Niçin Su? Miletli Tales’in Evren Anlayışı Bu makalenin yazılış amacı antik Greklerin felsefe öncesi hikmetlerinde su unsuru- nun var olduğunu göstermektir. Bu çalışma söz konusu unsuru gayr-i maddi olarak ele alınması gerektiğini öne sürmektedir. Thales’in felsefesini daha iyi anlaşılmasını sağla- mak için onun görüşleriyle ilgili bütün mevcut bilgileri bir araya getireceğiz

  5. Geochemical Processes and compacted bentonite FEBEX with a thermohydraulic gradient with a thermohydraulic gradient; Procesos geoquimicos y modificaciones texturales en bentonita FEBEX compactada sometida a un gradiente termohidraulico

    Energy Technology Data Exchange (ETDEWEB)

    Leguey Jimenez, S.; Cuevas Rodriguez, J.; Martin Barca, M.; Vigil de la Villa Mencia, R.; Ramirez Martin, S.; Garcia Gimenez, R. [Universidad Autonoma de Madrid (Spain)

    2002-07-01

    At present, the main source of High Level radioactive Waste (HLW) is the electrical energy production during all sep of developing. In almost all the countries with nuclear programs, the option for the final management of HLW is the Deep Geological Repository (DGR), based on the concept of multi barrier. According to this concept, the wastes is isolated from biosphere by the interposition of confinement barrier. In the context of an investigation of the near field for a repository of HLW, the FEBEX Project, a set of laboratory test has been designed to give a better understanding of the thermo-hydro-mechanical and geochemical behaviour of the compacted bentonite as a confinement barrier. The object of these work is to analyse the properties of the bentonite and its behaviour under conditions that will be found in a repository. The precipitation of mineral phases, due to local changes in the chemical equilibrium and the hydration itself, can produce changes in the salinity of the interstitial water and in the microstructural organisation of the clay particles. the hydraulic and mechanical properties of the bentonite can be modified by the special conditions of the barrier. (Author)

  6. Fast reactor: an experimental study of thermohydraulic processes in different operating regimes

    Science.gov (United States)

    Opanasenko, A. N.; Sorokin, A. P.; Zaryugin, D. G.; Trufanov, A. A.

    2017-05-01

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.

  7. CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, B.; Zhang, H.; Han, B.; Yang, B.W. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology; Mo, S.J.; Ren, H.B.; Qin, J.M.; Zuo, C.P. [China Nuclear Power Design Co. Ltd., ShenZhen (China)

    2016-07-15

    The integrity and thermal hydraulic characteristics of steam generator are of great concern in the nuclear industry. The tube support plates (TSP), one of the most important components of the steam generator, not only support the heat transfer tubes, but also affect the flow dynamic and thermal hydraulic characteristics of the secondary-side flow inside the steam generator. Different working conditions, ranging from single-phase adiabatic condition to two-phase high-void boiling condition, are simulated and analyzed. Calculated void fraction, under simple geometry, agrees well with the experiment data whilst the simulated heat transfer coefficient is tremendously close to the empirical correlation. Temperature, void fraction, and velocity distributions in different locations show reasonable distribution. The simulation results indicate that TSP can enhance the heat transfer in the secondary side of the steam generator. On the top of TSP, with the increase in cross-section flow area, the back-flow phenomenon occurs, which might lead to the contamination of precipitation.

  8. An investigation on thermo-hydraulic performance of a flat-plate channel with pyramidal protrusions

    NARCIS (Netherlands)

    Ebrahimi, Amin; Naranjani, Benyamin

    2016-01-01

    In this study, a flat-plate channel configured with pyramidal protrusions are numerically analysed for the first time. Simulations of laminar single-phase fluid flow and heat transfer characteristics are developed using a finite-volume approach under steady-state condition. Pure water is selected

  9. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  10. Thermo-hydraulics of the Peruvian accretionarycomplex at 12°S

    Science.gov (United States)

    Kukowski, Nina; Pecher, Ingo

    1999-02-01

    Coupled heat and fluid transport at the Peruvian convergent margin at 12°S wasstudied with finite element modelling. Structural information was available from two seismicreflection lines. Heat production in the oceanic plate, the metamorphic basement, and sedimentswas estimated from literature. Porosity, permeability, and thermal conductivity for the modelswere partly available from Ocean Drilling Program (ODP) Leg 112; otherwise we used empiricalrelations. Our models accounted for a possible permeability anisotropy. The decollement was bestmodelled as a highly permeable zone (10 -13 m 2). Permeabilities of thePeruvian accretionary wedge adopted from the model calculations fall within the range of 2 to7×10 -16 m 2 at the ocean bottom to a few 10 -18 m 2 at the base and need to be anisotropic. Fluid expulsion at the sea floor decreases graduallywith distance from the deformation front and is structure controlled. Small scale variations of heatflux reflected by fluctuations of BSR depths across major faults could be modelled assuming highpermeability in the faults which allow for efficient advective transport along those faults. The models were constrained by the thermal gradient obtained from the depth of bottomsimulating reflectors (BSRs) at the lower slope and some conventional measurements. We foundthat significant frictional heating is required to explain the observed strong landward increase ofheat flux. This is consistent with results from sandbox modelling which predict strong basalfriction at this margin. A significantly higher heat source is needed to match the observed thermalgradient in the southern line.

  11. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  12. Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Jakubec Jakub

    2017-04-01

    Full Text Available The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

  13. Study of thermohydraulic characteristics of upgraded feedwater collector in PGV-440 steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Tarankov, G.A.; Trunov, N.B.; Titov, V.F. [OKB Gidropress (Russian Federation); Urbansky, V.V. [Rovno NPP (Ukraine); Lenkei, I.; Notarosh, M. [Paks NPP (Hungary)

    1995-12-31

    Reconstruction of feedwater distribution collector was performed at unit 1 of Rowno NPP. Main results of measurements of temperatures in water volume, reparation characteristics and impurities distribution are presented. Analysis of tests results and design criteria is given. (orig.).

  14. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mendler, O J; Takeuchi, K; Young, M Y

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

  15. Thermohydraulics analysis for pipeline increase capacity; Estudo termohidraulico para ampliacao de capacidade de oleodutos

    Energy Technology Data Exchange (ETDEWEB)

    Carneiro, Leonardo Motta; Krause, Philipe Barroso; Pires, Luis Fernando G. [Pontificia Univ. Catolica do Rio de Janeiro (PUC-Rio), RJ (Brazil). Dept. de Engenharia Mecanica. Nucleo de Simulacao Termohidraulica de Dutos - SIMDUT; Souza, Antonio Geraldo de [TRANSPETRO, Rio de Janeiro, RJ (Brazil)

    2008-07-01

    This paper intent to assist the development of a oil pipeline expansion study. It will show that each pipeline has its own solution, because it has several variables of technical, economical and environmental order, as well as several ones of political nature. (author)

  16. Study of the thermohydraulics of CO2 discharge from a high pressure reservoir

    NARCIS (Netherlands)

    Ahmad, M.; Osch, M.B.V.; Buit, L.; Florisson, O.; Hulsbosch-Dam, C.; Spruijt, M.; Davolio, F.

    2013-01-01

    An experimental test set up has been constructed to carry out controlled CO2 release experiments from a high pressure vessel. The test set up is made up of a 500l stainless steel vessel where CO2 can be introduced up to high pressures and where controlled releases can be conducted. The work

  17. Experimental Study of Thermo-hydraulic Characteristics of Surfaces with In-line Dimple Arrangement

    Directory of Open Access Journals (Sweden)

    S. A. Burtsev

    2015-01-01

    Full Text Available The paper presents a conducted experimental study of the heat exchange intensification on the surfaces covered with a regular vortex-generating relief that is an in-line array of the shallow hemispherical dimples. Using 12 configuration options with the Reynolds numbers in the range of (0.2-7.0 106 as an example, it analyses how a longitudinal and cross step of the in-line dimple array (density dimples effects on the processes of heat exchange intensification and resistance.The monocomponent strain-gauge balance allows us to define a value of the resistance coefficient by direct weighing of models (located in parallel in a flow of "relief" and smooth "reference" ones being under study. Distribution fields of heat – transfer factor are determined by recording a cooling process of the surface of studied models having high spatial and temporary resolution. All researches were conducted with one-shot data record of these thermal and hydraulic measurements for the smooth (reference surfaces and the studied surfaces covered with a regular vortex-generating relief (dimples. The error of determined parameters was no more than ±5%.The oil-sooty method allows us to visualize flow around a regular relief and obtain a flow pattern for 12 options of dimples configuration. The analysis has been carried out and a compliance of the flow patterns with the field of heat-transfer factors has been obtained.It has been found that for the in-line configuration a Reynolds analogy factor for most models is nonlinearly dependent on the Reynolds number. The friction intensification, at first, falls (to some Reynolds number and, further, starts increasing, tending to the friction intensification value with self-similarity flow around. Thus with increasing Reynolds number, the heattransfer factor intensification falls (more slowly than resistance intensification.

  18. A thermo-hydraulic analysis of the superconducting proposal for the TF magnet system of FAST

    Energy Technology Data Exchange (ETDEWEB)

    Polli, G.M., E-mail: gianmario.polli@enea.it [EURATOM-ENEA, C.R. Frascati, Via E. Fermi, 45, IT-00044 Frascati, Rome (Italy); Corte, A. della; Di Zenobio, A.; Muzzi, L.; Reccia, L.; Turtu, S.; Brolatti, G.; Crisanti, F.; Cucchiaro, A.; Pizzuto, A.; Villari, R. [EURATOM-ENEA, C.R. Frascati, Via E. Fermi, 45, IT-00044 Frascati, Rome (Italy)

    2011-10-15

    FAST (Fusion Advanced Studies Torus), the Italian proposal of a satellite facility to ITER, is a compact tokamak (R{sub 0} = 1.82 m, a = 0.64 m, triangularity {delta} = 0.4) able to investigate non linear dynamics effects of {alpha}-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO. Currently, ENEA is investigating the feasibility of a superconducting solution for the magnet system. This paper focuses on the analysis of the TF magnets thermal behavior. In particular, utilizing only the room available in the resistive design and referring to one of the most severe scenario envisaged for FAST, the minimum temperature margin in the coil has been calculated for a thermal load distribution on winding and cable jacket due to nuclear heating only.

  19. Using statistical sensitivities for adaptation of a best-estimate thermo-hydraulic simulation model

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.c [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Dong Chuan Road 800, Shanghai 200240 (China); Kerner, A. [Institute for Energy Economy and Application Technology, Technical University of Munich, Walther-Meissner-Str. 2, 85748 Garching (Germany); Schaefer, A. [ISaR Institute for Safety and Reliability at Technical University of Munich, Walther-Meissner-Str. 2, 85748 Garching (Germany)

    2010-10-15

    On-line adaptation of best-estimate simulations of NPP behaviour to time-dependent measurement data can be used to insure that simulations performed in parallel to plant operation develop synchronously with the real plant behaviour even over extended periods of time. This opens a range of applications including operator support in non-standard-situations, improving diagnostics and validation of measurements in real plants or experimental facilities. A number of adaptation methods have been proposed and successfully applied to control problems. However, these methods are difficult to be applied to best-estimate thermal-hydraulic codes, such as TRACE and ATHLET, with their large nonlinear differential equation systems and sophisticated time integration techniques. This paper presents techniques to use statistical sensitivity measures to overcome those problems by reducing the number of parameters subject to adaptation. It describes how to identify the most significant parameters for adaptation and how this information can be used by combining: -decomposition techniques splitting the system into a small set of component parts with clearly defined interfaces where boundary conditions can be derived from the measurement data, -filtering techniques to insure that the time frame for adaptation is meaningful, -numerical sensitivities to find minimal error conditions. The suitability of combining those techniques is shown by application to an adaptive simulation of the PKL experiment.

  20. Modelization Post-test experiment IFA-650.10 HALDEN with FRAP series codes; Modelizacion post-test del experimento HALDEN IFA-650.10 con los codigos de la serie FRAP

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo, I.; Herranz, L. E.

    2013-07-01

    There is a need to review the criteria for security relating to LOCA accidents , including the effect of different materials of pod, as well as conditions of high burned as fuel. In this work is modeled with code FRAPTRAN-1.4 the IFA-650.10 experiment executed in the experimental reactor HALDEN. It is an approximation to the thermo-hydraulic rod-refrigerant and the results are compared with experimental measurements. The thermal behavior shows good agreement with the experimental measures; mechanical parameters are observed light quality deviations in pod and very good quantitative agreement in the maximum elongation; the diameter calculated at the end of the simulation above - predicts the post-irradiation values and oxide presents a good deal.

  1. Effects of RCP trip when recovering HPSI during LOCA in a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Montero-Mayorga, Javier, E-mail: fj.montero@alumnos.upm.es; Queral, César; Rivas-Lewicky, Julio; González-Cadelo, Juan

    2014-12-15

    Highlights: • If HPSI is recovered during SBLOCA and RCPs are tripped core damage can be reached. • If the RCPs are tripped once the accumulators have injected the damage can be avoided. • If only 2 out of 3 RCPs are tripped the damage can be also avoided. • Improvements are proposed to the EOPs in order to avoid possible damage. - Abstract: Current Westinghouse Emergency Operating Procedures (EOPs) indicate initially that the operator must keep the reactor coolant pumps (RCPs) running during a Small Break Loss of Coolant Accident (SBLOCA) if there is unavailability of high pressure safety injection (HPSI) system in order to cool the core by forced convection. However, the crew must follow different EOPs along the transient depending on its evolution. In these EOPs there are several conditions which indicate the necessity of tripping one or more RCPs when HPSI is recovered. In this paper the occurrence of a SBLOCA with unavailability of HPSI has been analyzed with a model of Almaraz Nuclear Power Plant (Westinghouse 3 Loop) for TRACE code V5.0 patch 1. Two different approaches have been considered: the first one, taking into account Optimal Recovery Guidelines (ORGs) and in the second approach, the transition to Function Restoration Guidelines (FRGs) due to inadequate core cooling (ICC) conditions is considered. Results of this paper lead to the implementation of an improvement in current EOPs regarding how many RCPs should be tripped during SBLOCA sequences.

  2. LoCa LoPa myelopathy: is prevention better than cure?

    Science.gov (United States)

    Pandita, Kamal Kishore; Razdan, Sushil; Pandita, Sarla

    2017-01-01

    Manifestations of primary hypoparathyroidism are produced by neuromuscular irritability or by extraosseous calcifications. We present a patient of primary hypoparathyroidism who had extensive calcification of brain parenchyma, and was suffering from chronic, generalised and progressive stiffness of body due to cervical compressive myelopathy, caused by calcification of posterior longitudinal ligament and ligamentum flavum. By presenting this case we wanted to emphasize the usefulness of meticulous clinical examination to differentiate the stiffness caused by myelopathy from that which is caused by possible coexisting extrapyramidal disorder. This case presentation also builds the hypothesis that early diagnosis and institution of early and appropriate treatment has potential to prevent the complications arising from extraosseous calcifications in patients with primary hypoparathyroidism.

  3. Política habitacional e locação social em Curitiba

    Directory of Open Access Journals (Sweden)

    Tomás Antonio Moreira

    Full Text Available A partir de dados censitários, o artigo analisa a política habitacional de interesse social na Região Metropolitana de Curitiba, entre 2000 e 1010. Os dados sobre os fluxos migratórios revelam que, embora essa região atraia maior número de migrantes no estado, já não é para a capital que se dirige a maioria. As cidades de São José dos Pinhais e Colombo, dotadas de menor infraestrutura, constituem o maior polo de atração. Com foco em habitação para famílias de baixa renda, o aluguel social em especial, são apresentadas as condições das unidades habitacionais, a infraestrutura disponível e os aspectos demográficos e socioeconômicos dos chefes de família. A leitura dos dados é contextualizada por uma recuperação histórica das políticas habitacionais promovidas pelo Estado desde a década de 1970.

  4. Analysis of fission product transport under terminated LOCA conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gieseke, J.A.; Baybutt, P.; Jordan, H.; Denning, R.S.; Wooton, R.O.

    1977-12-30

    An analytical model was developed to allow preditions of the source term to the containment as dependent on release from the fuel pins and deposition within the primary system. The model was developed into a flexible computer code adaptable to various geometrical arrangements and flow paths. The calculational framework was established in such a way to permit, in principle, the determination of particle and vapor transport and deposition in a general system of control volumes connected by fluid flow in an arbitrary way. This framework requires as input the rate of fission product release to the primary flow, as a function of time for each vapor and particulate species to be considered, and a complete, time dependent, thermal-hydraulic description of the system. TRAP-PWR, TRAP-BWR and TRACK computer codes are specializations, described in detail in the text of this report, of the general framework. The nature of these specializations depends strongly on the degree of detail with which the transporting fluid medium is modeled.

  5. La loca astucia de la voz del superyó en el imperativo capitalista del consumo

    Directory of Open Access Journals (Sweden)

    Megdy David Zawady

    2009-05-01

    Full Text Available Lacan inventa el aparato discursivocomo un modo de formalizar lasmodalidades de lo imposible propiasdel lazo social. Actualmente,la estabilización del discurso delamo en su estilo capitalista pone encuestión la articulación discursivamisma, al introducir una dinámicaque no está soportada en la imposibilidad.El discurso capitalistase sostiene en una forclusión de lacastración y su retorno en lo real esla circularidad de los imperativos degoce del superyó, cuyo ser vocalconvoca al sujeto a la realizaciónde la pulsión de muerte. En estecontexto la ciencia trabaja en laproducción de gadgets para la satisfacciónde un sujeto consumidorque al obedecer al imperativo, esconsumido por el sistema.Palabras clave: lazo social, real,capitalismo, superyó, discursoanalítico.

  6. Leading Learning and Teaching: An Exploration of "Loca"' Leadership in Academic Departments in the UK

    Science.gov (United States)

    Irving, Kate

    2015-01-01

    This paper reports on a small-scale longitudinal study of "local" leadership roles at two UK universities. The research explored perceptions of the leadership provided by a specific group of staff who held roles for enhancing learning and teaching. Based on ethnographic design principles, the study was based at one UK higher education…

  7. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

    1995-09-01

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

  8. Experimental investigation on the causes for pellet fragmentation under LOCA conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bianco, Andrea

    2015-07-23

    An experimental investigation was conducted in hot cells on single fuel rod segments to appraise the behavior of fuel pellets fragmentation during a loss of coolant accident in a light water reactor. In pursuing the conceptual design of the experiment, calculations were performed to study the thermal-hydraulics boundary conditions and the fuel rod behavior during the transient. The experiment's results encompass non-destructive and destructive examinations. In order to describe the resulting fuel fragments size distribution, a semi-empirical correlation was derived from the fractal theory.

  9. 'Locas' y 'fuertes': cuerpos precarios en el Guayaquil del siglo XXI

    National Research Council Canada - National Science Library

    Sancho Ordonez, Fernando

    2011-01-01

    ... explicar las formas como los cuerpos de quienes subvierten el sistema dominante de sexo y genero han sido excluidos del espacio publico de una ciudad que al empezar el siglo XXI inicio un proceso de 'regeneracion urbana...

  10. Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jae Won; Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1998-12-31

    Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous`s empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing. 6 refs., 8 figs. (Author)

  11. Always Running. La Vida Loca: Gang Days in L.A.

    Science.gov (United States)

    Rodriguez, Luis J.

    This autobiographical narrative describes the early life of Luis J. Rodriguez, a journalist and poet who was immersed in the youth gang culture of Los Angeles (California). Framed by the story of the pull of the gang life for the poet's son, it recounts his experiences from his childhood on the United States-Mexico border through his family's…

  12. Loca amoena in Don Quixote. The art of transition in pastoral episodes

    Directory of Open Access Journals (Sweden)

    Miguel Ángel Márquez

    2016-11-01

    Full Text Available Don Quixote’s pastoral episodes allow us to analyze Cervantes’s subtleness and irony in the art of transition. Cervantes intentionally uses the locus amoenus as a topos to link the pastoral episodes with the comic realistic story of Don Quixote and Sancho. The most striking feature of these meticulous transitions is that the locus amoenus, which would correspond to the pastoral stories, is transferred to the primary diegetic level, replacing the hegemonic chronotopos of the camino real. The role of this technique is to smooth the transition to the bucolic world. Moreover, this analysis helps us interpret the pastoral events as a tragic counterpoint (Marcela and Chrysostom, Leandra, and to perceive Don Quixote’s siesta (the “feigned Arcadia” as the starting point of his decline.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  14. Experimental study of thermo-hydraulic characteristics of natural circulation loop at water and FC-72 boiling under atmospheric pressure

    Science.gov (United States)

    Kaban’kov, O. N.; Sukomel, L. A.; Zubov, N. O.; Yagov, V. V.

    2017-10-01

    The results of experimental study of thermo and hydraulic characteristics of flow boiling of water and FC-72 in natural circulation loop under atmospheric pressure are presented. The experimental data have been obtained in the range of wall heat flux densities (6 – 70) kW/m2 for water and (4.6 – 30) kW/m2 for FC-72. These two liquids differ substantially in thermophysical properties so it makes it possible to extend the range of reduced pressures almost for an order of magnitude without changing the technical parameters of experimental setup. An additional information for the analysis of flow pattern influence on onset of instability and unstable circulation mechanism have been obtained as the result. The flow up tube of the loop had inner diameter 9.1 mm and consisted of two section – heated one 98 diameters length (that is 65 % of total tube length) and upper adiabatic section with length 48 diameters. Different circulation regimes were realized in experiments: mixed regimes with single phase and boiling zones in the heated part of the tube and boiling regimes along the full length of the heated section. The experimental data on circulation velocity (flow rate) and wall temperature distributions (including pulsating components of temperature and velocity) are presented in dependence on wall heat flux density and liquid subcooling at the inlet to the heated zone. At water experiments autooscillating regimes of boiling flows were observed within the whole range of inlet liquid subcoolings up to saturation temperature and at all wall heat flux densities from lowest one (10 kW/m2) to somewhat upper limiting value of 64 kW/m2. At higher heat fluxes the two-phase boiling flow was stable not only in saturation inlet liquid temperature but also at low subcoolings. In FC-72 experiments the flow was stable at all realized heat flux densities within the range of inlet liquid subcoolings (2 – 20) °C.

  15. Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H.; Kimura, N.; Miyakoshi, H. [Japan Nuclear Cycle Development Institute, Reactor Engineering Group, O-arai Engineering Center, Ibaraki (Japan); Nagasawa, K. [Nuclear Energy System Incorporation, O-arai Office, Ibaraki (Japan)

    2001-07-01

    Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

  16. Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR. Verification of the simulation results

    Energy Technology Data Exchange (ETDEWEB)

    Farahani, Aref Zarnooshe [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club

    2017-07-15

    Development of a steady-state model is the first step in nuclear safety analysis. The developed model should be qualitatively analyzed first, then a sensitivity analysis is required on the number of nodes for models of different systems to ensure the reliability of the obtained results. This contribution aims to show through sensitivity analysis, the independence of modeling results to the number of nodes in a qualified MELCOR model for a Westinghouse type pressurized power plant. For this purpose, and to minimize user error, the nuclear analysis software, SNAP, is employed. Different sensitivity cases were developed by modification of the existing model and refinement of the nodes for the simulated systems including steam generators, reactor coolant system and also reactor core and its connecting flow paths. By comparing the obtained results to those of the original model no significant difference is observed which is indicative of the model independence to the finer nodes.

  17. A three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yamano, Hidemasa [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, 4002 Narita, O-arai, Ibaraki 311-1393 (Japan)], E-mail: yamano.hidemasa@jaea.go.jp; Tobita, Yoshiharu; Fujita, Satoshi [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, 4002 Narita, O-arai, Ibaraki 311-1393 (Japan)

    2009-09-15

    The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core, including control rods, SIMMER-IV has been developed as a direct extension of SIMMER-III to three dimensions with retaining exactly the same physical models as SIMMER-III. Recently, the parallelization of SIMMER-IV has been achieved, allowing application to reactor calculations within available computational resources. A three-dimensional simulation using SIMMER-IV has drawn more realistic accident scenario including a late stage during the transition phase. Additional static neutronic calculations identified major factors significantly influencing the reactivity change shown in the SIMMER-IV simulation.

  18. Evaluation method for core thermohydraulics during natural circulation in fast reactors. Numerical predictions of inter-wrapper flow

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Nagasawa, Kazuyoshi [Nuclear Energy System Incorporation, Oarai Office, Oarai, Ibaraki (Japan)

    2002-08-01

    Decay heat removal using natural circulation is one of significant functions for a reactor. As the decay heat removal system, a direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this system, cold sodium is provided in an upper plenum of reactor vessel and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such phenomena was developed, which modeled each subassembly as a rectangular duct with gap region and also the upper plenum. This numerical simulation method was verified by a sodium test and also a water test. We applied this method to the natural circulation in a 600 MWe class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. (author)

  19. Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System

    National Research Council Canada - National Science Library

    Magdalena Nemś; Artur Nemś; Jacek Kasperski; Michał Pomorski

    2017-01-01

    ... to climate conditions in Poland. The system’s working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater...

  20. Nos reíamos como locas, como locas, como locas… La catástrofe eufórica en Yo era una chica moderna de César Aira.

    Directory of Open Access Journals (Sweden)

    Lucrecia Velasco Esquivel

    2013-07-01

    Full Text Available César Aira’s Yo era una chica moderna can be read as a metaphorical reconstruction of the social dynamics surrounding the argentine socio-economical crisis culminated at the end of 2001. The national catastrophe is portrayed with an euphoric tone that draws ironically on a literary tradition accustomed to associating popular advance with the monster figure and to focalizing the narrative through his violent point of view, thus communicating a sadistic glee in the very act of destruction. Nevertheless, the euphoria permeating the discourse can be also seen in the perspective of a new alliance between artists and lower classes emerging from the civil society’s spontaneous reaction to the crisis, specifically from a series of collective actions promoting horizontality, participation and solidarity between social groups. Aira’s novel seems to celebrate this unusual sensibility, especially in the literary field, where innovative practices began to diffuse as an answer to the new social demands.

  1. Fabrication Control Plan for ORNL RH-LOCA ATF Test Specimens to be Irradiated in the ATR

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Howard, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teague, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-06-01

    The purpose of this fabrication plan is (1) to summarize the design of a set of rodlets that will be fabricated and then irradiated in the Advanced Test Reactor (ATR) and (2) provide requirements for fabrication and acceptance criteria for inspections of the Light Water Reactor (LWR) – Accident Tolerant Fuels (ATF) rodlet components. The functional and operational (F&OR) requirements for the ATF program are identified in the ATF Test Plan. The scope of this document only covers fabrication and inspections of rodlet components detailed in drawings 604496 and 604497. It does not cover the assembly of these items to form a completed test irradiation assembly or the inspection of the final assembly, which will be included in a separate INL final test assembly specification/inspection document. The controls support the requirements that the test irradiations must be performed safely and that subsequent examinations must provide valid results.

  2. Sacra loca y armamento. Algunas reflexiones en torno a la presencia de armas no funcionales en contextos rituales

    Directory of Open Access Journals (Sweden)

    Gabaldón Martínez, María del Mar

    2010-12-01

    Full Text Available Many weapons from the Ancient Past have been found in archaeological contexts defined by its ritual character such as graves, sanctuaries or votive deposits. Precisely from these archaeological contexts came most of the findings of weapons with no military function. Some of them were manipulated (symbolically or physically for ritual purposes and the others were made for an exclusively ritual or ceremonial function.Un gran número de armas del pasado proceden de contextos arqueológicos definidos por su carácter ritual, ya sean las sepulturas, los santuarios o los «depósitos votivos». En estos contextos arqueológicos se han hallado la mayor parte de las armas que no tienen función para el combate, bien porque han sido transformadas (física o simbólicamente, bien porque han sido creadas (destinadas para tener un uso exclusivamente ritual o ceremonial. [fr] Un grand nombre d’armes antiques provient de contextes archéologiques définis par leur caractère rituel, qu’il s’agisse des sépultures, des sanctuaires ou des dépôts votifs. C’est dans ces contextes qu’on été retrouvées la majorité des armes dépourvues de fonctionnalité militaire, soit parce que leurs caractéristiques ont été altérées (matériellement ou symboliquement, soit parce qu’elles ont été spécialement fabriquées dans un but rituel ou cérémonial.

  3. Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

    Directory of Open Access Journals (Sweden)

    Eltayeb Yousif

    2017-01-01

    Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

  4. Sensitivity Analyses in Small Break LOCA with HPI-Failure: Effect of Break-Size in Secondary-Side Depressurization

    Science.gov (United States)

    Kinoshita, Ikuo; Torige, Toshihide; Yamada, Minoru

    2014-06-01

    In the case of total failure of the high pressure injection (HPI) system following small break loss of coolant accident (SBLOCA) in pressurized water reactor (PWR), the break size is so small that the primary system does not depressurize to the accumulator (ACC) injection pressure before the core is uncovered extensively. Therefore, steam generator (SG) secondary-side depressurization is necessary as an accident management in order to grant accumulator system actuation and core reflood. A thermal-hydraulic analysis using RELAP5/MOD3 was made on SBLOCA with HPI-failure for Oi Units 3/4 operated by Kansai Electoric Power Co., which are conventional 4 loop PWR plants. The effectiveness of SG secondary-side depressurization procedure was investigated for the real plant design and operational characteristics. The sensitivity analyses using RELAP5/MOD3.2 showed that the accident management was effective for a wide range of break sizes, various orientations and positions. The critical break can be 3 inch cold-leg bottom break.

  5. Mechanistic modeling of pool film-boiling and quench on a Candu calandria tube following a critical break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T.; Luxat, J.C. [McMaster University, A315 JHE Building, 1280 Main St.W. Hamilton, ON, L8S 4L7 (Canada)

    2008-07-01

    Following a postulated critical LBLOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a Candu CT (approximately 13 cm). A second order, non-linear and non-homogeneous ODE for vapour film thickness has been derived. The variation of steady state vapour film thickness prior to quench as a function of subcooling temperature, wall superheat, and incident heat flux is examined. The CT outer surface heatup rate and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (authors)

  6. LocaTag - An NFC-based system enhancing instant messaging tools with real-time user location

    OpenAIRE

    Köbler, Felix; Koene, Philip; Krcmar, Helmut; Altmann, Matthias; Leimeister, Jan Marco

    2010-01-01

    In recent years instant messaging tools have been successfully introduced to support communication and collaboration processes in work environments. Research suggests that the use of instant messaging tools lead to an increased feeling of connectedness, social presence and awareness within a collaborative group. Consequently, the goal of our research is to improve communication and collaboration within a group through automated and real-time dissemination of presence information, using instan...

  7. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  8. 77 FR 40823 - Airworthiness Directives; Airbus Airplanes

    Science.gov (United States)

    2012-07-11

    ... . For Thales Avionics service information identified in this proposed AD, contact Thales Avionics... Include Other AOA Probes Airbus stated that certain Thales AOA probes, part number (P/N) C16291AB, have been modified in accordance with Thales Avionics Service Bulletin C16291A-34-007, August 27, 2009...

  9. HYDRA-I: a three-dimensional finite difference code for calculating the thermohydraulic performance of a fuel assembly contained within a canister

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.

    1980-12-01

    A finite difference computer code, named HYDRA-I, has been developed to simulate the three-dimensional performance of a spent fuel assembly contained within a cylindrical canister. The code accounts for the coupled heat transfer modes of conduction, convection, and radiation and permits spatially varying boundary conditions, thermophysical properties, and power generation rates. This document is intended as a manual for potential users of HYDRA-I. A brief discussion of the governing equations, the solution technique, and a detailed description of how to set up and execute a problem are presented. HYDRA-I is designed for operation on a CDC 7600 computer. An appendix is included that summarizes approximately two dozen different cases that have been examined. The cases encompass variations in fuel assembly and canister configurations, power generation rates, filler materials, and gases. The results presented show maximum and various local temperatures and heat fluxes illustrating the changing importance of the three heat transfer modes. Finally, the need for comparison with experimental data is emphasized as an aid in code verification although the limited data available indicate excellent agreement.

  10. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  11. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte I

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R.; Ahnert, C.; Naudin, A. E.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-07-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  12. Mal de las vacas locas: su influencia en la obtención del antígeno principal de Haemophilus influenzae tipo b

    OpenAIRE

    Jonatan Hernández-Roche; Osmir Cabrera Blanco; Maribel Cuello-Pérez; Luis Riverón Martínez; Arturo Talavera Coronel; Ana H. Callís-Díaz; Gustavo Sierra-González

    2007-01-01

    Se evaluaron los parámetros fundamentales del proceso de obtención del polisacárido capsular de Haemophilus influenzae tipo b tales como: medio de cultivo, tipo de hemina empleada como factor de crecimiento y la expresión de cápsula polisacarídica por diferentes cepas. Los medios de cultivo evaluados fueron: infusión cerebro corazón, Mueller Hinton y una variante de Frantz modificado. Las heminas que fueron evaluadas procedieron de las casas comerciales siguientes: Fluka (origen bovino), Sigm...

  13. An improvement of estimation method of source term to the environment for interfacing system LOCA for typical PWR using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seok Jung; Kim, Tae Woon; Ahn, Kwang Il [Risk and Environmental Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

  14. IMPACT OF THE CHEMICAL FORM OF IN-CONTAINMENT SOURCE ON FISSION PRODUCT RELEASE FROM WWER-1000/V-320 TYPE NPP CONTAINMENT DURING LOCA

    Directory of Open Access Journals (Sweden)

    Adam Kecek

    2016-12-01

    Full Text Available Nuclear power plant accidents may be followed by a release of fission products into the environment. This release is dependent on several phenomena, such as chemistry, pressure, type of the accident etc. The aim of this paper is to assess the impact of the chemical form of iodine on the fission product release into the environment.

  15. Estudio de la composición fenóloca del aceite de oliva virgen extra : caracterización y reactividad antioxidante

    OpenAIRE

    Becerra Herrera, Mercedes

    2013-01-01

    La Tesis se ha basado en el estudio de la actividad antioxidante de compuestos fenólicos presentes en el aceite y la caracterización de los mismos. Así pues, en la primera investigación realizada se hace uso del fluoróforo 2,3-diazabiciclo[2.2.2]oct-2-eno (DBO) como alternativa al tradicional método espectrofotométrico en el cual se utiliza el radical libre 2,2-difenil-1-picrilhidrazil (DPPH•). Posteriormente se realizó una comparación de tres métodos para la extracción de fenoles del A...

  16. TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.

  17. Church ladies, good girls, and locas: Stigma and the intersection of gender, ethnicity, mental illness, and sexuality in relation to HIV risk

    OpenAIRE

    Collins, Pamela Y.; von Unger, Hella; Armbrister, Adria

    2008-01-01

    Inner city women with severe mental illness may carry multiple stigmatized statuses. In some contexts these include having a mental illness, being a member of an ethnic minority group, being an immigrant, being poor, and being a woman who does not live up to gendered expectations. These potentially stigmatizing identities influence both the way women’s sexuality is viewed and their risk for HIV infection. This qualitative study applies the concept of intersectionality to facilitate understand...

  18. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

    2017-08-15

    An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  19. 75 FR 19203 - Airworthiness Directives; Bombardier, Inc. Model CL-600-2B19 (Regional Jet Series 100 & 440...

    Science.gov (United States)

    2010-04-14

    ...) airplanes, serial numbers 7003 and subsequent, certificated in any category, that are equipped with Thales... December 1, 2008, references Thales Avionics Service Bulletins 45150340-31- 004 and C16258A-27-002, both...

  20. 75 FR 38947 - Airworthiness Directives; Airbus Model A330-200 and A330-300 Series Airplanes, and Model A340-200...

    Science.gov (United States)

    2010-07-07

    ... describes the unsafe condition as: Investigation conducted by Thales on probes revealed oil residue between...://www.airbus.com . For Thales Avionics service information identified in this proposed AD, contact Thales--Aerospace Division, 105, avenue du General Eisenhower--BP 63647, 31036 Toulouse Cedex 1, France...

  1. 77 FR 10693 - Airworthiness Directives; Airbus Airplanes

    Science.gov (United States)

    2012-02-23

    [email protected] ; Internet http://www.airbus.com . For Thales Avionics service information identified in this proposed AD, contact Thales Avionics, Retrofit Manager, 105, Avenue du G n ral Eisenhower... Thales on the removed probes revealed oil residue between the stator and the rotor parts of the AoA vane...

  2. Security-aware Virtual Machine Allocation in the Cloud: A Game Theoretic Approach

    Science.gov (United States)

    2015-01-13

    Security Achilles Heel." www.zdnet.com. CBS Interactive, 29 Mar. 2014. Web. 14 July 2014. [12] " Thales Finds Organizations More Confident Transferring...Sensitive Data to the Cloud despite Data Protection Concerns." Https://www.thales esecurity.com/company/press/news. Thales E Security, 25 June 2013

  3. 75 FR 68698 - Airworthiness Directives; Airbus Model A330-201, -202, -203, -223, -223F, -243, and -243F...

    Science.gov (United States)

    2010-11-09

    ... aviation product. The MCAI describes the unsafe condition as: * * * * * Investigation conducted by Thales... conducted by Thales on the removed probes revealed oil residue between the stator and the rotor parts of the... inspection of the Thales Avionics AoA probe P/N [part number] C16291AA in order to identify the suspect parts...

  4. 78 FR 19085 - Airworthiness Directives; Airbus Airplanes

    Science.gov (United States)

    2013-03-29

    ... confirmed by flight data analysis. Investigation conducted by Airbus and Thales on the removed probes... number (s/n) of each installed Thales Avionics Part Number (P/N) C16291AA AoA probe and the replacement... 2012-0236R1, dated December 17, 2012. In addition, Airbus stated that Thales Avionics has issued...

  5. Joint Tactical Radio System Handheld, Manpack, and Small Form Fit Radios (JTRS HMS)

    Science.gov (United States)

    2015-12-01

    User Objective System (Navy managed waveform). The RR Full and Open Competition contract was awarded on April 29, 2015 to two vendors, Thales Group and...Contracts Contract Identification Appropriation: Procurement Contract Name: Thales - Rifleman Radio FOC Contractor: Thales Defense & Security

  6. Development of data acquisition system for test circuit for the Thermo-Hydraulic Laboratory of CDTN; Desenvolvimento de sistema de aquisicao de dados para circuito de testes do Laboratorio de Termo-Hidraulica do CDTN

    Energy Technology Data Exchange (ETDEWEB)

    Corrade, Thales Jose Rodrigues; Mesquita, Amir Zacarias; Santos, Andre Augusto Campagnole dos, E-mail: thalescorrade@hotmail.com, E-mail: amir@cdtn.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2013-07-01

    The Circuit Water-Air (CWA), present in the Laboratorio de Termo-Hidraulica of the Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear (CDTN / CNEN), has been used to evaluate devices present in nuclear fuel elements of a PWR (Pressurized Water Reactor). Currently, a segment of 5x5 beam simulators grids with spacer bars is being tested, serving one of the activities under the Project FUJB / FINEP / INB - 'Development of New Generation of Nuclear Fuel Element '. For the measurements of pressure drop along this beam, a system of data acquisition based on Basic language was created. Although this system is efficient and robust, their resources are very limited. Therefore, it was decided to use the software LabVIEW® implementing a more versatile and modern system. This article describes the new data acquisition system, and presents some results. The main parameters are monitored: temperature, density, dynamic viscosity, Reynolds number. The values of standard deviation, mean and uncertainty of an arbitrary channel are calculated. The system was installed and tested in the circuit under experimental conditions and showed satisfactory results.

  7. Obtaining of a simulator of thermohydraulics transients of the Cofrentes NPP using SNAP-TRACE platform; Obtencion de un simulador de transitorios termohidraulicos de la CN Cofrentes utilizando la plaraforma SNAP-TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Escriva, A.; Munoz-cobo, J. L.; Concejal, A.; Soler, A.; Melara, J.; Albendea, M.

    2013-07-01

    The model developed using point kinetics and simulate various transients interactively, such as the firing of feed water turbo-pumps or the closing of the valves of main steam (MSIVs). Developed models allow to visualize, through different screens, the behavior of the whole plant as well as its control system.

  8. Silicon microchannel cooling panel for NA62 Giga-Tracker, proposal G.Nuessle : a first thermo-hydraulic layout attempt for use with monophase, liquid C6F14 circulation

    CERN Document Server

    Wertelaers, P

    2010-01-01

    In this proposal, where the hydraulic regime (laminar) of the liquid (monophase) is simple, analytical recipes can be worked out. They show clearly the scaling laws in the relation from coolant pressure budget to panel temperature chart. If the line length is irreducible, then the individual channels cannot become arbitrarily small, even if, then, there can be many to take the total thermal load. The reason is that the "capacitive" component would explode. Apart from showing this, the Note also discusses cross-coupling effects between adjacent U-shaped channels.

  9. Application of COBAYA3 code to the multiscale analysis with neutronic-thermohydraulic connection in PWR reactors; Aplicacion del codigo COBAY A3 al analisis multiescala con acoplamiento neutronico-termohidraulico en PWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J.; Herrero, J. J.; Lozano, J. A.; Aragones, J. M.; Cuervo, D.; Garcia-Herranz, N.; Ahnert, C.

    2010-07-01

    This paper presents the application of COBAYA3 codes in PWR reactors cores analysis. The methodology for the breakdown in subcommands by means of alternated dissections has been implemented in the above-mentioned code system as part of the work done by the first two authors in their doctoral thesis.

  10. DIFERENÇA Entre Características na Formação do Preço de Venda e Locação de Imóveis na Cidade de Vitória/es

    OpenAIRE

    LOPES, C. I.

    2015-01-01

    Essa dissertação estuda as características relevantes na formação do preço de venda e aluguel, analisando também as diferenças entre esses atributos para apartamentos na cidade de Vitória/ES, preenchendo uma lacuna ainda não desenvolvida, tendo em vista a possibilidade de comparação entre preços de aluguel e venda. O constructo teórico teve como fundamento abordagem de preços hedônicos, aplicada em estudos de Waugh (1928) e Court (1939), mas formalmente desenvolvida teoricament...

  11. La loca Ð El Gado: Un análisis de la figura del héroe de la novela El gato de sí mismo, de Uriel Quesada

    Directory of Open Access Journals (Sweden)

    José Pablo Rojas González

    2014-11-01

    Full Text Available El presente trabajo busca estudiar –a partir de los aportes teóricos de Mijaíl Bajtín– la figura del héroe de la novela El gato de sí mismo de Uriel Quesada. Se pretende caracterizar al héroe de este texto costarricense como una nueva variante del “soñador” y del “hombre del subsuelo” bajtinianos. Esta caracterización permitirá señalar la importancia de atentar contra el estereotipo, de atentar contra el «orden sexual» establecido, a favor –siempre– de la libertad del ser humano

  12. Conceptual design of safety injection tanks using saturated water

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Min; Jeong, Yong Hoon; Chang, Won Joon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2009-07-01

    Safety Injection Tanks (SITs) which is the one of Safety Injection System (SIS) play an important role in mitigating the Loss of Coolant Accidents (LOCAs) in Pressurized Water Reactor (PWR). APR1400 has the advanced 4 SITs directly connected to a reactor vessel. We expect the capacity of the SITs is getting more important since the coolant from SITs equipped with a FD during LBLOCA can replace the injection from low pressure safety injection pumps (LPSIPs). In designing a larger capacity SIT, we may have three problems; the excessively large volume for pressurized N{sub 2} gas, which is about 1/3 of the total volume, the difficulties controlling injection flowrate and the solubility of the non-condensable N{sub 2} gas in the coolant. In here, there is the contradiction which is 'there must be nitrogen gas for pressurization but there must not be nitrogen gas for more coolant.' For this problem, the axiomatic design (AD) theory enabled us to define or regularize the intrinsic problem which is termed the coupling and the contradiction. TRIZ facilitates creating solutions on the contradiction. This study proposes a conceptual design of SITs which are pressurized by steam from the saturated water as a demonstration of the conceptual design framework, AD theory and TRIZ. The purpose of this conceptual design is to increase coolant volume and to reduce N{sub 2} gas volume in SITs. In order to investigate the feasibility of the proposed design, we derived an analytical model to find the heat loss of saturated water and thermo-hydraulic safety analysis using MARS3.1. To confirm the safety and integrity of core, we conducted LBLOCA simulation to find peak cladding temperature (PCT) of design using the proposed SITs comparing with the conventional SITs. From the analysis results, the benefits of the new SIT design were observed in terms of the PCT, the quenching time and the size. And the new SIT design may enable emergency core cooling water to be injected

  13. Dicty_cDB: CFE675 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CD825453 |CD825453.1 BN25.060N03F011129...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 AF076243 |AF076243.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 AV567523 |AV567523.1 Arabidopsis thaliana

  14. Dicty_cDB: VFG884 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 8e-06 1 BX827174 |BX827174.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 8e-06 1 AF076243 |AF076243.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 8e-06 1 dna update 2004. 2.19 Homology vs Protein

  15. Dicty_cDB: VFH157 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 64 3e-12 3 BX831354 |BX831354.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 64 4e-12 3 AY045584 |AY045584.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 64 4e-12 3 BG525546 |BG525546.1 51-5 Stevia field

  16. Dicty_cDB: VFF755 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 1e-05 1 BX827174 |BX827174.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 1e-05 1 AF076243 |AF076243.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 1e-05 1 AV567523 |AV567523.1 Arabidopsis thaliana

  17. Dicty_cDB: VFH291 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 1e-05 1 AF076243 |AF076243.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 1e-05 1 BX829292 |BX829292.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 1e-05 1 AV567523 |AV567523.1 Arabidopsis thaliana

  18. Dicty_cDB: CFB725 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 64 7e-13 3 BX831354 |BX831354.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 64 7e-13 3 AY045584 |AY045584.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 64 8e-13 3 BG525546 |BG525546.1 51-5 Stevia field

  19. Dicty_cDB: VFJ441 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 BX827174 |BX827174.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 BX829292 |BX829292.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 AV567523 |AV567523.1 Arabidopsis thaliana

  20. Dicty_cDB: VFH125 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 64 2e-07 2 BX831354 |BX831354.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 64 2e-07 2 AY045584 |AY045584.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 64 2e-07 2 CA523816 |CA523816.1 KS12029A03 KS12

  1. Design of a Multi-Touch Tabletop for Simulation-Based Training

    Science.gov (United States)

    2014-06-01

    gmail.com Directorate of Land Synthetic Environments, CFB Kingston Kingston, ON, Canada Liam Porter liam.porter@ca.thalesgroup.com Thales ...Porter liam.porter@ca.thalesgroup.com Thales Canada Defense and Security Centre Kingston 945 Princess St. Kingston, ON, K7L 3N6, Canada Abstract... Thales Canada and the Command and Staff Training and Capability Development Center. We would like to gratefully acknowledge the contributions of

  2. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  3. Violence In Honduras: An Analysis of the Failure in Public Security and the States Response to Criminality

    Science.gov (United States)

    2014-06-01

    brings researchers. One of the promising films and research was by filmmaker Christian Poveda, but his 2008 documentary La Vida Loca126 portraying the...American Politics and Society 54, no. 1 (March 2012): 69. 126 Cristian Poveda, “ La Vida Loca,” accessed January 10, 2013, http//www.youtube.com/watch?v... La Vida Loca. The nexus of urbanization, an unusually large percentage of youth population, the decay of the typical nuclear family, and the

  4. Dicty_cDB: CFF884 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CD825453 |CD825453.1 BN25.060N03F011129...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 AL161500 |AL161500.2 Arabidopsis thaliana

  5. Dicty_cDB: VFG862 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CB264640 |CB264640.1 48-E014661-035-002-P12-T7R...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CD825453 |CD825453.1 BN25.060N03F011129

  6. Dicty_cDB: VFG464 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 BX827174 |BX827174.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 dna update 2004. 3. 9 Homology vs Protein

  7. Dicty_cDB: VFF806 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CB264640 |CB264640.1 48-E014661-035-002-P12-T7R...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CD825453 |CD825453.1 BN25.060N03F011129

  8. Dicty_cDB: VFF807 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CD825453 |CD825453.1 BN25.060N03F011129...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 AL161500 |AL161500.2 Arabidopsis thaliana

  9. Dicty_cDB: VFG726 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 AV567523 |AV567523.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 CB264640 |CB264640.1 48-E014661-035-002-P12-T7R

  10. Dicty_cDB: CFC781 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 AL765461 |AL765461.1 Arabidopsis thaliana...of strain col-0 of Arabidopsis thaliana (thale cress). 62 2e-05 1 dna update 2004. 3. 2 Homology vs Protein

  11. 12om Methodology: Process v1.1

    Science.gov (United States)

    2014-03-31

    12om Methodology: Process v1.1 Prepared by: Thales Canada, Defence and Security 1405 boul. du Parc-Technologique Québec...have the approval or endorsement of the Department of National Defence of Canada. Thales Document Control Number (DCN): 2014C.005-REP-03-AT5 Rev. 01

  12. Cruise Missile Penaid Nonproliferation: Hindering the Spread of Countermeasures Against Cruise Missile Defenses

    Science.gov (United States)

    2014-01-01

    flare/chaff dispensers SOURCES: (Left) Thales promotional image; (right) W. Vinten Ltd. promotional image. Penaids of various types may need to be...As of July 28, 2014: http://www.rand.org/pubs/research_reports/RR378.html “ Thales Vicon 78 Series 500,” Jane’s Unmanned Aerial Vehicles and Targets

  13. Subchannel Code Benchmarking to Columbia University 4x4 and Pacific Northwest Laboratory 2x6 Bundle Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Kang Hoon; Oezdemir, Erdal; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    The subchannel code is used to assess the safety of a reactor core at the steady-state and transient conditions. KEPCO Nuclear Fuel (KNF) has been developed new subchannel code, THALES, for PWR core design application. In this study, we are comparing the THALES result with VIPRE-01 code result utilizing bundle test data. VIPRE-01 was developed under EPRI sponsorship and has been used by U.S. PWR commercial nuclear utilities, historically. THALES and VIPRE-01 codes were benchmarked to two kind of bundle test data which were at the steady-state and transient conditions. THALES predicted fluid velocity and temperature profile of bundle test data well and the error rate between THALES and VIPRE-01 was very small.

  14. Performance Assessment in a Heat Exchanger Tube with Opposite/Parallel Wing Twisted Tapes

    Directory of Open Access Journals (Sweden)

    S. Eiamsa-ard

    2015-02-01

    Full Text Available The thermohydraulic performance in a tube containing a modified twisted tape with alternate-axes and wing arrangements is reported. This work aims to investigate the effects of wing arrangements (opposite (O and parallel (P wings at different wing shapes (triangle (Tri, rectangular (Rec, and trapezoidal (Tra wings and on the thermohydraulic performance characteristics. The obtained results show that wing twisted tapes with all wing shape arrangements (O-Tri/O-Rec/O-Tra/P-Tri/P-Rec/P-Tra give superior thermohydraulic performance and heat transfer rate to the typical twisted tape. In addition, the tapes with opposite wing arrangement of O-Tra, O-Rec, and O-Tri give superior thermohydraulic performances to those with parallel wing arrangement of P-Tra, P-Rec, and P-Tri around 2.7%, 3.5%, and 3.2%, respectively.

  15. Thermal–hydraulic characteristics for CANFLEX fuel channel using burnable poison in CANDU reactor

    National Research Council Canada - National Science Library

    Bae, Jun Ho; Jeong, Jong Yeob

    2015-01-01

    ... (natural uranium) fuel channels. The distributions of fuel temperature and critical channel power for the CANFLEX-BP fuel channel are calculated using the NUclear Heat Transport CIRcuit Thermohydraulics Analysis Code (NUCIRC...

  16. Analysis of thermohydraulics fluctuations in NPP Trillo with basis/PARCSv2.7. Validation code and comparison with results of SIMULATE-3K; Analisis de fluctuaciones termohidraulicas en C.N. Trillo con RELAP5/PARCSv2.7. Validacion del codigo y comparacion con resultados de SIMULATE-3K

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fenoll, M.; Barrachina, T.; Miro, R.; Verdu, G.; Bermejo, J. A.; Lopez, A.; Ortega, A.

    2013-07-01

    In this work, a RELAP5/PARCSv2.7 model of TRILLO NPP core and the obtainment of INCORE and EXCORE detectors signals is presented. For its validation, Control Rod drop transient real data is used. Besides, the results are compared with SIMULATE-3K results obtained by CNAT. Different transients triggered by moderator temperature perturbations at the core inlet are performed, and the results are compared with SIMULATE-3K results for these transients.

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

    Energy Technology Data Exchange (ETDEWEB)

    Salazar C, J.H.; Nunez C, A. [CNSNS, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D.F. (Mexico); Chavez M, C. [UNAM, Facultad de Ingenieria, DEPFI Campus Morelos (Mexico)]. E-mail: hsalazar22@prodigy.net.mx

    2004-07-01

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

  18. Propagation and Interactions of Ultrahigh Power Light: Relativistic Nonlinear Optics

    Science.gov (United States)

    2014-09-30

    THALES LASER. It is a CPA laser system [2], seeded by an ultra-short broadband nano-joule oscillator (SYNERGY, FEMTOLASERS, Inc.) at 76 MHz with pulse...laser (JADE, THALES LASER). An electro-optic pulse cleaner in this amplifier stage filters the amplified spontaneous emission (ASE) and pre-pulses to...Glass lasers (ATLAS PLUS, THALES LASER) each producing 25 J at 527 nm with a repetition rate of 0.1 Hz. The beam profiles of all the 25-J pump lasers are

  19. Taxonomic determination of five groups of slugs and life cycle studies of the prevalent group in a commercial crop of alstroemeria in the loca lit y of Madrid, Cundinamarca Determinación taxonómica de cinco grupos de babosas y estudio del ciclo de vida del grupo predominante en un cultivo comercial de alstroemeria de Madrid - Cundinamarca

    Directory of Open Access Journals (Sweden)

    Martínez. John Wilson

    1994-06-01

    Full Text Available Slugs are an important pest due to the wide range of crops that they are able to attack. Lately, they have been registe red as an important pest in alstroemeria crop, where they leed with the tender buds arthe plant base, avoiding new productive stems development and causing yield decreases.
    Due to the lack of  knowledge about those molluscs and the high soil moisture and low light conditions into alstroemeria crop that enhance slugs development as a pest, this research was done. This work that was carried out in Jardines de
    Colombia Ltda (Madrid-Cundinamarca under laboratory and comercial greeenhouse conditions includes taxonomic determination and lile cycle study of the predominant slug group. Also, lour other groups that are present in the same crop were taxonomically determined. Results show that egg
    has an incubation time 01 17.17 days under laboratory and 21.7 days under greenhouse conditions. Slug lile cycle has a 124.5 and 123.9 days duration, respectively. Samples of each slug group sended to specialists, be long to Limacidae and Arionidae lamilies, and Deroceras and Limax genera were present in Limacidae lamily. Genera in Arionidae lamily weren't determined.Las babosas son una plaga importante debido a el amplio rango de cultivos que pueden llegar a atacar. Ultimamente, se han registrado como plaga importante en el cultivo de Alstroemeria en donde se alimentan de los brotes tiernos de la base de la planta, evitando el desarrollo de nuevos tallos
    productivos y, por consiguiente, originando disminución en el rendimiento. Debido a la falta de conocimiento sobre estos moluscos y a las condiciones de alta humedad y baja luminosidad dentro del cultivo de alstroemeria, que favorecen
    su desarrollo, se hizo indispensable la realización de la presente investigación. El trabajo fué realizado en la Empresa Jardines de Colombia Ltda (Madrid- Cundinamarca, tanto bajo condiciones de laboratorio, como de invernadero comercial. Incluye la determinación taxonómica y el estudio del ciclo de vida del grupo de babosas predominante.
    También, se realizó la determinación taxonómica de otros cuatro grupos presentes en el mismo cultivo. Los resultados indican que el huevo presenta un tiempo de incubación de 17,17 dias en el laboratorio y de 21,7 dias en invernadero. El
    ciclo de vida tiene una duración de 124,5 y 123,9 dias, respectivamente. Muestras de cada uno de los grupos de babosas, enviadas a especialistas, pertenecen a las familias Limacidae y Arionidae, presentandose los géneros Deroceras y Limax en Limacidae. Los géneros de la familia Arionidae no
    fueron determinados.

  20. O espólio da encosta do Lado Ocidental do Castelo de Alcácer do Sal (LOCAS Alentejo, Portugal : a terra sigillata de tipo itálico decorada e marcas de oleiro II (Um projecto de João Carlos Faria = Samian ware of the western slope of the Castle of...

    Directory of Open Access Journals (Sweden)

    Eurico de Sepúlveda

    2014-12-01

    Full Text Available Com este artigo dá-se início ao segundo ciclo e final do estudo sobre as cerâmicas exumadas no Lado Ocidental do Castelo de Alcácer do Sal. Privilegiámos a terra sigillata de tipo itálico na sua vertente decorada, para além das várias marcas de oleiro. Dedicamos esta comunicação ao nosso grande Amigo, arqueólogo, coordenador e impulsionador do estudo deste espólio, o JOÃO CARLOS FARIA, que infelizmente não o pôde terminar.Con este artículo se da inicio al segundo ciclo y final del estudio de la cerámica fina localizada en el sitio arqueológico que se encuentra ubicado en la cuesta occidental del castillo de la ciudad de Alcácer do Sal (Alentejo, Portugal. Nuestra investigación tuvo como objeto de estudio una parte residual de la terra sigillata de origen itálica, decorada, así como sus marcas de alfarero. Dedicamos este artículo a nuestro gran Amigo arqueólogo y coordinador de este proyecto, JOÃO CARLOS FARIA, quien infelizmente no pudo terminarlo.The authors present the ultimate results of a study concerning Italian terra sigillata excavated on the western slope of the castle of Alcácer do Sal. Decorated sherds and fragments with potter’s stamps were analysed. This paper is in honour of our beloved friend who suddenly died in 2007.

  1. Verification of ceramic structures

    NARCIS (Netherlands)

    Behar-Lafenetre, S.; Cornillon, L.; Rancurel, M.; Graaf, D. de; Hartmann, P.; Coe, G.; Laine, B.

    2012-01-01

    In the framework of the "Mechanical Design and Verification Methodologies for Ceramic Structures" contract [1] awarded by ESA, Thales Alenia Space has investigated literature and practices in affiliated industries to propose a methodological guideline for verification of ceramic spacecraft and

  2. The Utility of Open Source Software in Military Systems

    National Research Council Canada - National Science Library

    Esperon, Agustin I; Munoz, Jose P; Tanneau, Jean M

    2005-01-01

    .... The companies involved were THALES and GMV. The MILOS project aimed to demonstrate benefits of Open Source Software in large software based military systems, by casting off constraints inherent to traditional proprietary COTS and by taking...

  3. Lääne firmad peavad Eesti radari pärast kohtulahingut / Toomas Mattson

    Index Scriptorium Estoniae

    Mattson, Toomas, 1970-

    2001-01-01

    Radarihankekonkursil osalenud Itaalia radaritootja Alenia Marconi Systems ja Prantsuse firma Thales vaidlustasid Tallinna halduskohtus valitsuse sajalase korralduse, mille alusel sõlmis kaitseministeerium radari ostmise lepingu USA firmaga Lockheed-Martin

  4. RAAK MKB Wireless Sensortechnologie bij calamiteiten : Werkpakket 3 Proximity

    NARCIS (Netherlands)

    Leeuwen, van H. (Henk); Sondervan, N. (Niels)

    2011-01-01

    Het project Wireless Sensor Technologie bij Calamiteiten is een samenwerkingsverband tussen Saxion, Thales Nederland (de dochterondernemingen D-CIS Lab en Iseti), Ambient Systems, Ti-WMC, het beveiligingsbedrijf Vigilat, het Regionaal Centrum Criminaliteitspreventie en Veiligheidsregio’s

  5. Uudised : Mozarti tundmatu ooper Berliinis. Kuurort avab poe. Muda Music - lõpuks ometi!

    Index Scriptorium Estoniae

    2006-01-01

    W. A. Mozarti koomilise ooperi "L'oca de Cairo" ("Kairo hani") esietendusest 15. apr. Deutsche Operis Berliinis. Plaadifirma Kuurort Recordsi internetikauplusest Kuurortshop (www. kuurortshop.com). Eesti plaadifirmasid koondavast netisaidist mudamusic.com

  6. Vehicle kinematics in turns and the role of cornering lamps in driver vision.

    Science.gov (United States)

    2010-11-01

    "SAE Recommended Practice J852 and ECE Regulations 119 and 48 for cornering lamps : were compared. Photometric points described in each specification were then compared : to naturalistic low-speed turn trajectories produced by 87 drivers. Future loca...

  7. Protein (Cyanobacteria): 302754 [PGDBj - Ortholog DB

    Lifescience Database Archive (English)

    Full Text Available lisation of periplasmic protein complexes Dactylococcopsis salina PCC 8305 MIVVGFAL...YP_007170462.1 1117:5821 1118:178 13034:352 292566:352 13035:352 putative periplasmic protein (DUF2233)/Loca

  8. Safety for Older Consumers: Home Safety Checklist

    Science.gov (United States)

    ... ground fault circuit interrupters, or GFCIs, in potentially damp locations such as the kitchen, bathroom, garage, near ... ground fault circuit interrupters, or GFCIs, in potentially damp loca- tions such as the kitchen, bathroom, garage, ...

  9. 77 FR 38094 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

    Science.gov (United States)

    2012-06-26

    ... met during a TS-allowed period of radiation monitor inoperability. There are no design changes to the... Large Break LOCA Analysis,'' Revision 1, that implements AREVA's NRC-approved topical report, EMF- 2103...

  10. Radiation degradation of plastic insulating materials

    Energy Technology Data Exchange (ETDEWEB)

    Bartonicek, B.; Hnat, V.; Janovsky, I.; Pejsa, R. [Nuclear Research Institute, Rez (Czech Republic)

    1995-10-01

    Several types of polymeric compounds, used as insulating and sheathing materials of cables, were subjected to accelerated thermal and radiation ageing and to LOCA test. The stability of materials was evaluated via their mechanical properties, namely strain at break. (Author).

  11. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

  12. HPLWR fine mesh core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Temesvari, E.; Maraczy, C.; Hegyi, G.; Hordosy, G.; Molnar, A. [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research

    2014-08-15

    The European version of Supercritical Water Reactors (SCWR), the High Performance Light Water Reactor (HPLWR) operates in the thermodynamically supercritical region of water. Our basic objective was to elaborate a stationary coupled neutronic-thermohydraulic code capable for the calculation of the actual 3-pass core design with fuel assembly clusters. The calculations covered the neutronic transport calculations of HPLWR fuel assemblies, the coupled neutronic-thermohydraulic global calculations and the pin-wise analysis. Applying conservative assumptions, the relation to the linear heat rate and maximum cladding temperature limits was checked for the equilibrium cycle of HPLWR with this new code system.

  13. Numerical simulation of two phase flows in heat exchangers; Simulation numerique des ecoulements diphasiques dans les echangeurs

    Energy Technology Data Exchange (ETDEWEB)

    Grandotto Biettoli, M

    2006-04-15

    The report presents globally the works done by the author in the thermohydraulic applied to nuclear reactors flows. It presents the studies done to the numerical simulation of the two phase flows in the steam generators and a finite element method to compute these flows. (author)

  14. Post-test analysis of the experiment 5.2C - total loss of feed water at the BETHSY test facility

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E.; Schaefer, F.

    1998-10-01

    The BETHSY-test facility is a 1:100 scaled thermohydraulic model of a 900 MW(el) pressurized water reactor (FRAMATOME). The test facility is mainly designed to investigate various accident scenarios and to provide an experimental data base for code validation and for the verification of accident management measures. (orig.)

  15. Analysis code for medium and small rupture accidents in ATR. LOTRAC/HEATUP

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    In the evaluation of thermo-hydraulic and fuel temperature transient changes in the events which are classified in medium and small rupture accidents of reactor coolant loss that is the safety evaluation event of the ATR, the analysis code for synthetic thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC and the detailed analysis code for fuel temperature HEATUP are used, respectively. By using the LOTAC, the thermo-hydraulic behavior of reactor cooling facility and the temperature behavior of fuel at the time of blow-down are analyzed, and also the characteristics of changing reactor thermal output is analyzed, considering the functioning characteristics of emergency core cooling system. Based on the data of thermo-hydraulic behavior obtained by the LOTRAC, the time of beginning the turn-around of fuel cladding tube temperature obtained by the data of ECCS pouring characteristics, the heat transfer rate after the turn-around and so on, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The LOTRAC code, the HEATUP code, various analysis models, and rupture simulation experiment are reported. (K.I.)

  16. Simplified 3D model of a PWR reactor vessel using fluid dynamics code ANSYS CFX computational; Modelo simplificado 3D de la vasija de un reactor PWR mediante el codigo de dinamica de fluidos computacional ANSYS CFX

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, M.; Miro, R.; Barrachina, T.; Verdu, G.

    2011-07-01

    This paper presents the results from the calculation of the steady state simulation with model of CFD (computational fluid dynamic) operating under conditions of operation at full power (Hot Full Power). Development and the CFD model results show the usefulness of these codes for calculating 3D of the variable thermohydraulics of these reactors.

  17. Numerical investigation of the High Temperature Reactor (VHTR) using computational fluid dynamics; Investigacao numerica do Reator de Alta Temperatura (VHTR) utilizando fluidodinamica computacional

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Joao Pedro C.T.A.; Santos, Andre A. Campagnole dos; Mesquita, Amir Z., E-mail: jpctap@cdtn.br, E-mail: aacs@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG),Belo Horizonte, MG (Brazil). Lab. de Termo-Hidraulica

    2013-07-01

    This work consists to evaluate and continue the study that is being developed in the Laboratory of Thermo-Hydraulics of the CNEN/CDTN (Centro de Desenvolvimento da Tecnologia Nuclear), aiming to validate the methods and procedures used in the numerical calculations of fluid flow in fuel elements of the core of the VHTR.

  18. Heat transfer and phenomenology in severe accidents in spent fuel pools with MAAP5; Transmision de calor y fenomenologia en accidentes severes en piscinas de combustible gastado con MAAP5

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz, J. A.; Gil, E.; Uruburu, A.; Rey, P.

    2013-07-01

    The code Thermo-hydraulic MAAP5 includes in their latest versions a module that allows you to analyze the evolution of an accident occurring in the pool of spent fuel from a nuclear power plant in their latest versions. This module is a preliminary version and there is interest from stations and reference centres in Spain to know in depth its capabilities.

  19. on coupled system of navier-stokes equations and temperature in ...

    African Journals Online (AJOL)

    Dr. Anthony Peter

    ABSTRACT. This paper deals with the coupled system of Navier-Stokes equations and temperature (Thermohydraulics) in a strip in the class of spatially non-decaying (infinite-energy) solutions belonging to the properly chosen uniformly local Sobolev spaces. The global well-posedness and dissipativity of the Navier- ...

  20. Development of the Approach by States method and thermodynamical study of a 1300 MWe PWR type reactor following a complete water loss of vapor generator alimentation with the Cathare 2 code; Developpement de la conduite APE et etude thermohydraulique d'un REP 1300 MWe suite a un accident de perte totale d'eau alimentaire des generateurs de vapeur avec le code Cathare 2

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, F

    1998-06-30

    The objective of this report is to study the thermohydraulic behavior of a 1300 MWe PWR type reactor for a complete loss accident in water supplying of vapor generators. The Cathare computer code has been used in this aim. (N.C.)

  1. Proceedings of the 1984 American Nuclear Society Midwest Student Conference. Our energy future

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    Sixty-seven abstracts are included, grouped under the following session headings: systems analysis, industrial development, and economics; miscellaneous; reactor safety; thermohydraulics, gas dynamics, and MHD; fusion technology and plasma physics; radiation dosimetry, data reduction, and medical imaging; instrumentation; and neutronics. (DLC)

  2. New Design Heaters Using Tubes Finned by Deforming Cutting Method

    Science.gov (United States)

    Zubkov, N. N.; Nikitenko, S. M.; Nikitenko, M. S.

    2017-10-01

    The article describes the results of research aimed at selecting and assigning technological processing parameters for obtaining outer fins of heat-exchange tubes by the deformational cutting method, for use in a new design of industrial water-air heaters. The thermohydraulic results of comparative engineering tests of new and standard design air-heaters are presented.

  3. Procedure of coupling of the codes TRAC-BFI Y PARCSv2.7; Procedimiento de acoplamiento de los codigos TRAC-BF1 y PARCSv2.7

    Energy Technology Data Exchange (ETDEWEB)

    Jambrina, A.; Barrachina, T.; Miro, R.; Verdu, G.; Concejal, A.; Melara, J.

    2012-07-01

    With the aim of increasing the ability of the TRAC-BF1 code analysis, has coupled with the neutronic code 3D PARCS v2.7. Thus, there are coupled code neutronic thermohydraulic that allows the simulation of transient neutron 3D and processes hydraulic multi-channel geometry 1D.

  4. Advanced ATHLET model for the UPTF facility

    Energy Technology Data Exchange (ETDEWEB)

    Nikonov, S. [RRC “Kurchatov Inst.” (Russian Federation); Pasichnyk, I.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

    2011-07-01

    The aim of the presented study is to determine the limits of application of the one dimensional system code ATHLET by describing in a pseudo 3D manner the thermohydraulic modelling of a reactor pressure vessel. For this purpose ATHLET simulation results are compared with available experimental data of the UPTF facility. Developed is a detailed ATHLET model based on a special preprocessor which is able to generate automatically a three dimensional nodalization of the facility. Based on this nodalization the preprocessor creates the necessary ATHLET input data set, including a description of the thermohydraulic objects and the connections between them. To take into account a spatial structure of interacting objects, advanced capabilities of ATHLET system code, namely cross connections in a system of parallel thermohydraulic channels are applied. Moreover in a new version of ATHLET several improvements are introduced: 1) an algorithm of the ATHLET Jacobian preconditioner is optimized; 2) a new direct sparse matrix solver based on the KLU algorithm of Tim Davis is implemented which gives a substantial acceleration of calculations for a large number of control volumes and makes feasible transient simulations with a detailed (ten thousands of control volumes) description of the investigated facility. The obtained comparisons have proved the advantages of the new modelling and applied computational algorithms and have confirmed the ability of the ATHLET code to describe in a pseudo 3D manner the thermohydraulic processes in a reactor pressure vessel. (author)

  5. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  6. Trend and Health Management Analyzer (THEMA): A Step Forward to an Integrated Analysis Environment

    Science.gov (United States)

    Piras, Annamaria; Masiello, Stefano; Fasano, Giorgio

    2014-08-01

    This note focuses on the fundamental subject of trend analysis and health management, in space engineering. Thales Alenia Space has been involved in this demanding issue with increasing commitment, with the objective of taking advantage from previous experience, investigating the state of the systems currently in use and forecasting their behavior in the long term. THEMA (Trend and HEalth Management Analyzer), an advanced analysis environment, has been developed by Thales Alenia Space for this purpose. It is described hereinafter, pointing out the operational context it has originated from, its main features and benefits. Insights on further development are provided.

  7. Simulation of in-reactor experiments with the ELOCA.Mk5 code. AECL research No. AECL-11133

    Energy Technology Data Exchange (ETDEWEB)

    Klein, M.E.; Arimescu, V.I.; Carlucci, L.N.

    1994-12-31

    ELOCA.Mk5 is a FORTRAN-77 computer code developed to model the thermo-mechanical response and associated fission-product release behavior of CANDU fuel elements during high-temperature transients such as large-break loss of coolant accidents (LOCA). This paper reports the results of model runs conducted to simulate two in-reactor LOCA experiments, using ELOCA.Mk5 in the Mk4S mode. Mk4S is a thermo-mechanical mode capable of performing a multi-segment analysis of a CANDU fuel element, accounting for axial variations in sheath temperatures, metallurgical regions, and reactor neutron flux. The first LOCA experiment consisted of four elements subjected to a coolant depressurization in the Power Burst Facility at Idaho Falls. The second consisted of a single fresh element, with an artificially set internal gas pressure, subjected to a coolant depressurization in the NRX reactor.

  8. CFD analysis of blockage length on a partially blocked fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Gabriel [Centro Universitário FEI (UNIFEI), São Paulo, SP (Brazil). Dept. de Engenharia Mecânica; Angelo, Edvaldo, E-mail: nikolas.scuro@gmail.com, E-mail: delvonei@ipen.br, E-mail: gangelo@fei.edu.br, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, São Paulo, SP (Brazil). Escola da Engenharia. Grupo de Simulação Numérica

    2017-07-01

    In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

  9. Development Considerations of AREVA NP Inc.'s Realistic LBLOCA Analysis Methodology

    Directory of Open Access Journals (Sweden)

    Robert P. Martin

    2008-01-01

    Full Text Available The AREVA NP Inc. realistic large-break loss-of-coolant-accident (LOCA analysis methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. This methodology conforms to the code scaling, applicability, and uncertainty (CSAU methodology developed by the Technical Program Group for the United States Nuclear Regulatory Commission in the late 1980s. In addition, several practical considerations were revealed with the move to a production application. This paper describes the methodology development within the CSAU framework and utility objectives, lessons learned, and insight about current LOCA issues.

  10. Evaluation of SPACE code for simulation of reactor scram due to unplanned loss of RCP power in OPR 1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seyun; Lee, Dong Hyuk; Kim, Yo Han; Ha, Sang Jun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) has been developing by KHNP with the cooperation with KEPCO E and C and KAERI. SPACE code is expected to be applied to the safety analysis for LOCA (Loss of Coolant Accident) and Non LOCA scenarios. SPACE code solves two fluid, three field governing equations and programmed with C++ computer language using object oriented concepts. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, the reactor trip accident due to the loss of RCP power in OPR1000 (Ulchin unit 4) was simulated with SPACE code.

  11. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  12. 78 FR 59295 - Airworthiness Directives; Airbus Airplanes

    Science.gov (United States)

    2013-09-26

    ... number (S/N) of each installed Thales Avionics (formerly SEXTANT), Part Number (P/N) C16291AA AoA sensor...) and (i) of this AD, provided that no AoA sensor has been replaced since first flight. (k) Parts...

  13. A Decision Analysis Framework for Evaluation of Helmet Mounted Display Alternatives for Fighter Aircraft

    Science.gov (United States)

    2014-12-26

    flown on F-16, AV-8B, Tornado 1990s TopSight/TopNight Thales Avionics Fielded on Mirage and Rafale 1990s Joint Helmet Mounted Cuing System (JHMCS...reliability, a decreasing categorical value function can be used since there is no intermediate gradation between location types (Figure 27). 84 Figure

  14. 76 FR 72996 - Eleventh Meeting: RTCA Special Committee 223 Airport Surface Wireless Communications

    Science.gov (United States)

    2011-11-28

    ... Leadership Designated Federal Official (DFO): Mr. Brent Phillips Co-Chair: Mr. Aloke Roy, Honeywell... Control December 7, 2011 MOPS WG Breakout Session Discussion of Security Sub-layer--Honeywell Review draft of Environmental (DO-160G)--Rockwell Collins Review draft PICS--EUROCAE (Thales) Review draft CSRL...

  15. The Origins of Science

    Indian Academy of Sciences (India)

    questioning, before the secrets of nature are yielded to us. Socrates said, "The unexamined life is not worth living." From Thales to Anaximander. Thus, this newly emerging tradition encouraged criticism, even of one's masters. We find that the second of the great Ionian philosophers, Anaximander, also of Miletus and both ...

  16. Pythagoras of Samos (c. 580-c. 500 BC)

    Science.gov (United States)

    Murdin, P.

    2000-11-01

    Mathematician, born in Samos, Ionia, taught by Thales and Anaximander on Miletus, founded a philosophical and religious school in Croton (in southern Italy) whose members, as well as having various beliefs that we would recognize as religious, believed that at its deepest level, reality is mathematical in nature. Pythagoras observed that vibrating strings produce harmonious tones when the ratios ...

  17. Aristarchus of Samos the ancient Copernicus

    CERN Document Server

    Heath, Sir Thomas

    2004-01-01

    This classic work traces Aristarchus of Samos's anticipation by two millennia of Copernicus's revolutionary theory of the orbital motion of the earth. Heath's history of astronomy ranges from Homer and Hesiod to Aristarchus and includes quotes from numerous thinkers, compilers, and scholasticists from Thales and Anaximander through Pythagoras, Plato, Aristotle, and Heraclides. 34 figures.

  18. Asymmetry of the two-beam geometry in EMCD experiments.

    Science.gov (United States)

    Rusz, J; Oppeneer, P M; Lidbaum, H; Rubino, S; Leifer, K

    2010-03-01

    We analyse theoretically the influence of the asymmetry of the two-beam geometry on quantitative measurements of the energy-loss magnetic chiral dichroism. Our simulations indicate that this asymmetry is not very strong inside or close to the Thales circle, but in other regions of the diffraction plane it can hinder an accurate extraction of the orbital to spin moment ratio.

  19. TOD characterization of the gatekeeper electro optical security system

    NARCIS (Netherlands)

    Gosselink, G.A.B.; Anbeek, H.; Bijl, P.; Hogervorst, M.A.

    2013-01-01

    The Triangle Orientation Discrimination (TOD) test method was applied to characterize thermal and visual range performance of the Gatekeeper Electro Optical Security System. Gatekeeper developed by Thales Nederland BV, is currently in use with the Royal Netherlands Navy. The system houses uncooled

  20. New high power CW klystrons at TED

    CERN Document Server

    Beunas, A; Marchesin, R

    2003-01-01

    Thales Electron Devices (TED) has been awarded a contract by CERN to develop and produce 20 units of the klystrons needed to feed the Large Hadrons Collider (LHC). Each of these delivers 300 kW of CW RF power at 400 MHz. Three klystrons have been delivered to CERN up to now.

  1. Resonance – Journal of Science Education | Indian Academy of ...

    Indian Academy of Sciences (India)

    Gangan Prathap. Articles written in Resonance – Journal of Science Education. Volume 1 Issue 3 March 1996 pp 3-3 Article-in-a-Box. Fermat and the Minimum Principle · Gangan Prathap · More Details Fulltext PDF. Volume 1 Issue 4 April 1996 pp 67-73 Reflections. The Origins of Science Thales' Leap · Gangan Prathap.

  2. Resonance – Journal of Science Education | Indian Academy of ...

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education. Pradipkumar H Keskar. Articles written in Resonance – Journal of Science Education. Volume 17 Issue 5 May 2012 pp 476-486 General Article. Multiplication: From Thales to Lie · Pradipkumar H Keskar · More Details Fulltext PDF ...

  3. Bildung--Then and Now in Danish High School and University Teaching and How to Integrate Bildung into Modern University Teaching

    Science.gov (United States)

    Olesen, Mogens Noergaard

    2010-01-01

    In the history of mankind three important philosophical and scientific revolutions have taken place. The first of these revolutions was the mathematical-axiomatic revolution in ancient Greece, when the philosophers from Thales of Miletus to Archimedes built up the abstract deductive method used in pure mathematics. The second took place in the…

  4. Resonance – Journal of Science Education | Indian Academy of ...

    Indian Academy of Sciences (India)

    Multiplication: From Thales to Lie · Pradipkumar H Keskar · More Details Fulltext PDF. pp 487-492 General Article. What is Transit? B S Shylaja · More Details Fulltext PDF. pp 493-496 Classroom. On a Definite Integral of the Fractional Part Function · Koundinya Vajjha · More Details Fulltext PDF. pp 497-504 Classroom.

  5. Interval availability analysis of a two-echelon, multi-item system

    NARCIS (Netherlands)

    Al Hanbali, Ahmad; van der Heijden, Matthijs C.

    2013-01-01

    In this paper we analyze the interval availability of a two-echelon, multi-item spare part inventory system. We consider a scenario inspired by a situation that we encountered at Thales Netherlands, a manufacturer of naval sensors and naval command and control systems. Modeling the complete system

  6. Design for undex survivability of an integrated modular mast

    NARCIS (Netherlands)

    Aanhold, H. van; Pel, S.; Bosman, T.

    2008-01-01

    Thales Naval Netherlands is currently developing a so-called Integrated Modular Mast (IMM), thereby supported by the Royal Netherlands Navy and TNO. There is a need for knowing the response of this IMM to underwater shock in such a way, that this can be used in the design of shock resistant sensor

  7. Experimental design of natural and accellerated bone and wood ageing

    DEFF Research Database (Denmark)

    Facorellis, Y.; Pournou, A.; Richter, Jane

    2015-01-01

    This paper presents the experimental design for natural and accelerated ageing of bone and wood samples found in museum conditions that was conceived as part of the INVENVORG (Thales Research Funding Program – NRSF) investigating the effects of the environmental factors on natural organic materials....

  8. Evolution: Yesterday, Today, Tomorrow

    Science.gov (United States)

    Mayer, William V.

    1977-01-01

    The author reviews research on the origins of life, beginning with Thales (636 B.C.), synthesized by C. Darwin in "The Origin of Species," continued by H. DeVries' mutation theory, and enhanced by the discovery in 1944 of DNA. For journal availability, see SO 505 260. (AV)

  9. Die Kosmologie der Griechen.

    Science.gov (United States)

    Mittelstraß, J.

    Contents: 1. Mythische Eier. 2. Thales-Welten. 3. "Alles ist voller Götter". 4. Griechische Astronomie. 5. "Rettung der Phänomene". 6. Aristotelische Kosmololgie. 7. Aristoteles-Welt und Platon-Welt. 8. Noch einmal: die Göttlichkeit der Welt. 9. Griechischer Idealismus.

  10. Interval availability analysis of a two-echelon, multi-item system

    NARCIS (Netherlands)

    Al Hanbali, Ahmad; van der Heijden, Matthijs C.

    2011-01-01

    In this paper we analyze the interval availability of a two-echelon, multi-item spare part inventory system. We consider a scenario inspired by a situation that we encountered at Thales Netherlands, a manufacturer of naval sensors and naval command and control systems. Modeling the complete system

  11. Dawn of Science

    Indian Academy of Sciences (India)

    SERIES ARTICLE. Dawn of Science. 1. The First Tottering Steps. T Padmanabhan. Keywords. Pyramids, Imhotep, Moscow. Papyrus, Thales of Miletus. In mankind's quest for knowledge, spanning the last four ... We believe it is the former and that the ancient literary fables cannot be taken too seriously in determining the ...

  12. Treatment Integrity of Elaborated Semantic Feature Analysis Aphasia Therapy Delivered in Individual and Group Settings

    Science.gov (United States)

    Kladouchou, Vasiliki; Papathanasiou, Ilias; Efstratiadou, Eva A.; Christaki, Vasiliki; Hilari, Katerina

    2017-01-01

    Background & Aims: This study ran within the framework of the Thales Aphasia Project that investigated the efficacy of elaborated semantic feature analysis (ESFA). We evaluated the treatment integrity (TI) of ESFA, i.e., the degree to which therapists implemented treatment as intended by the treatment protocol, in two different formats:…

  13. Correspondencia.

    OpenAIRE

    Vladimir Zaninovic

    2009-01-01

    Muy importantes y oportunas las publicaciones en la prensa internacional relacionadas con las �vacas locas y los pollos tóxicos�. Infortunadamente, en Colombia las personas creen que ese es un problema meramente europeo. Esta comunicación se refiere a la �locura y a la pobreza de libre importación

  14. Laser pulse heating of nuclear fuels for simulation of reactor power ...

    Indian Academy of Sciences (India)

    It is important to study the behaviour of nuclear fuels under transient heating conditions from the point of view of nuclear safety. To simulate the transient heating conditions occurring in the known reactor accidents like loss of coolant accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating system is under ...

  15. Laser pulse heating of nuclear fuels for simulation of reactor power ...

    Indian Academy of Sciences (India)

    Abstract. It is important to study the behaviour of nuclear fuels under transient heating conditions from the point of view of nuclear safety. To simulate the transient heating conditions occurring in the known reactor accidents like loss of coolant accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating ...

  16. 78 FR 38078 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

    Science.gov (United States)

    2013-06-25

    ... energy] release input for the LOCA containment response analysis. The [P a ] remains below the... accident previously evaluated? Response: No. The use of the non seismic Boric Acid Recovery System (BARS... describes a different method for submitting comments on a specific subject): Federal Rulemaking Web site: Go...

  17. Disease: H01310 [KEGG MEDICUS

    Lifescience Database Archive (English)

    Full Text Available cores on muscle biopsy and clinical features of a congenital myopathy. It is morphologically defined by loca...nning to a limited extent along the longitudinal axis of muscle fiber (minicores). Marked clinical variabili... respiratory impairment, whereas mutations in the RYR1 gene have been associated with a wider range of clinic

  18. Variations in levels of care within a hospital provided to acute ...

    African Journals Online (AJOL)

    5 to theatre. Conclusion. Significant variations exist in the level of obser- vations of vital signs between different geographical loca- tions within the hospital. This is problematic ... porters unaccompanied by medical staff in 3 cases. One patient .... must be completed and stuck onto the patient's file would force staff to formally ...

  19. 75 FR 32509 - Notice Applications and Amendments to Facility Operating Licenses Involving Proposed No...

    Science.gov (United States)

    2010-06-08

    ... Commission (NRC)-approved topical report (TR) EMF-2103(P)(A), Revision 0, ``Realistic Large-Break LOCA [Loss... Shearon Harris Nuclear Power Plant, Unit 1 safety analyses. TR EMF-2103(P)(A), Revision 0, was approved by..., failure mechanisms, or limiting single failures are introduced [that create a new or different accident...

  20. Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert W. [FPoliSolutions LLC, Murrysville, PA (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [FPoliSolutions LLC, Murrysville, PA (United States); Yurko, Joseph P. [FPoliSolutions LLC, Murrysville, PA (United States); Swindlehurst, Gregg [GS Nuclear Consulting, Charlotte, NC (United States); Zoino, Angelo [Univ. of Rome Tor Vergata (Italy)

    2015-09-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.

  1. STATUS OF WOOD PROCESSING AND STORAGE IN NIGERIA

    African Journals Online (AJOL)

    ES Obe

    Execution. 3. 3.1. Some Wood Available in Nigeria and their Local Names. There are different kinds of wood avail- able in Nigeria, irrespective of their loca- tion and usage. They include: Apa, Black. Afara, White Afara, Afromosia, Gmelina,. Idigbo, Canarium, Okan, Antiaria, Lagos. Mahogany, Berlinia, Ekki, Scented Guarea,.

  2. combination Dictionary

    African Journals Online (AJOL)

    rbr

    5. mad cow disease enfermedad de las vacas locas/encelofatía espongiforme bovina. Excuses abound: world markets have collapsed, diet-conscious Europeans are eating less red meat, some people in Britain fear it will give them mad cow dis- ease. (i). Usage notes are highlighted in light blue, and range from pragmatic ...

  3. Fully-Plastic Crack Growth in Asymmetric Plane Strain Bending.

    Science.gov (United States)

    1986-07-31

    initiation was to produce enogh plasticA -.. strain to initiate the crack without fatiquin9 the ligament. In, nort cas-2-, the structure is subject to...siinificartiv imp,:,rt- art and should be considered to avoid possible catastrophic 42ll PS f I. ! fa ilIur es. Loca! cr3ck growth ductility. In

  4. University and place branding: The case of universities located in ECC (European Capital of Culture) cities

    OpenAIRE

    Gábor Rekettye; Gyöngyi Pozsgai

    2015-01-01

    In the globalising landscape of higher education more and more universities are going international. These universities are facing growing competition, especially in enrolling international students. International competition forces them to use marketing and especially branding activity. University branding requires that the higher education institutions clearly define their differentiating features. One of the most important differentiating features is the place where the institution is loca...

  5. A 10-Hz terawatt class Ti:sapphire laser system: Development and ...

    Indian Academy of Sciences (India)

    875–881. A 10-Hz terawatt class Ti:sapphire laser system: Development and applications ... petawatt class CPA laser systems using different gain media have been realized. In particular, Ti:sapphire CPA laser .... video cameras to monitor the input and amplified output beams at different loca- tions and to correct their spatial ...

  6. Activation tagging of the LEAFY PETIOLE gene affects leaf petiole development in Arabidopsis thaliana

    DEFF Research Database (Denmark)

    van der Graaff, Eric; Dulk-Ras, A D; Hooykaas, P J

    2000-01-01

    In a screen for leaf developmental mutants we have isolated an activator T-DNA-tagged mutant that produces leaves without a petiole. In addition to that leafy petiole phenotype this lettuce (let) mutant shows aberrant inflorescence branching and silique shape. The LEAFY PETIOLE (LEP) gene is loca...

  7. Identifying Challenges and Opportunities for Residents in Upernavik as Oil Companies are Making a First Entrance into Baffin Bay

    DEFF Research Database (Denmark)

    Merrild, Anne; Tejsner, Pelle

    2016-01-01

    The oil industry is making its first entrance offshore in Baffin Bay in a time where Inuit residents on the northwest coast of Greenland are struggling to uphold a traditional way of living. The operating oil companies are encouraged by the Government of Greenland to promote a high degree of loca...

  8. ASSESSMENT OF SV ELEVATION MASK ON THE ACCURACY AND ACCESSIBILITY OF CERTAIN AIRCRAFT POSITIONING ON FINAL APPROACH WITH THE SIGNALS OF GLONASS

    OpenAIRE

    G. V. Krinitsky; A. V. Zimina; A. S. Zimin

    2015-01-01

    The dependence of the accuracy and availability of aircraft location on final approach on signals of GLONASS with the certain satellites elevation angle is determined. Estimation of the accuracy and accessibility of the aircraft loca-tion was calculated for angles of elevation of the satellites within the specified range for the various aircraft location.

  9. ASSESSMENT OF SV ELEVATION MASK ON THE ACCURACY AND ACCESSIBILITY OF CERTAIN AIRCRAFT POSITIONING ON FINAL APPROACH WITH THE SIGNALS OF GLONASS

    Directory of Open Access Journals (Sweden)

    G. V. Krinitsky

    2015-01-01

    Full Text Available The dependence of the accuracy and availability of aircraft location on final approach on signals of GLONASS with the certain satellites elevation angle is determined. Estimation of the accuracy and accessibility of the aircraft loca-tion was calculated for angles of elevation of the satellites within the specified range for the various aircraft location.

  10. Nigerian Journal of Paediatrics 2013;40(2)

    African Journals Online (AJOL)

    owner

    2012-09-01

    Sep 1, 2012 ... of the apex beat in childhood particularly with respect to age. Further, it may be interesting to know how the loca .... recruited from day care centres and nursery schools while children aged between 6 and 10 years were re- ..... 2007; 62-95. 11. Robinson TN, Dietz WH. Weight. Gain: Overeating to Obesity. In:.

  11. Sorghum allelopathy – from ecosystem to molecule

    Science.gov (United States)

    Sorghum allelopathy has been reported in a series of field experiments following sorghum establishment. In recent years, sorghum phytotoxicity and allelopathic interference have also been well-described in greenhouse and laboratory settings. Observations of allelopathy have occurred in diverse loca...

  12. Department of Energy's team's analyses of Soviet designed VVERs

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document provides Appendices A thru K of this report. The topics discussed respectively are: radiation induced embrittlement and annealing of reactor pressure vessel steels; loss of coolant accident blowdown analyses; LOCA blowdown response analyses; non-seismic structural response analyses; seismic analyses; S'' seal integrity; reactor transient analyses; fire protection; aircraft impacts; and boric acid induced corrosion. (FI).

  13. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  14. Investigating Constraint-Based Approaches for the Development of Agile Plans

    Science.gov (United States)

    2014-06-01

    as chosen du s. Three var ctivity. The rted positio o at the bou rrespondin intensity o all contribut n the type o erms of loca In TPEM...Company Cbt Tm will have to clear two routes, namely DIAM along A log team Plato Route know 4.2. The f synch are p tasks interv availa links

  15. Author Details

    African Journals Online (AJOL)

    Ikeobi, CON. Vol 1, No 1 (1999) - Articles Egg quality characteristics of four loca poultry species in Nigeria Abstract · Vol 1, No 1 (1999) - Articles Presence of the polydactyly gene in the Nigerian local chicken. Abstract · Vol 1, No 2 (1999) - Articles Bovine wastages in Abbatoir and slaughter slabs of Oyo State, Nigeria: ...

  16. RELIABILITY of FUEL ASSEMBLY EFFLUENT TEMPERATURES UNDER L0CA/LOPA CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Sachs, A.D.

    1999-06-21

    The purpose of this study was to ascertain whether or not the K-Reactor safety computers could calculate primarily false positive, but also false negative, and ''on-scale'' misleading fuel assembly average effluent temperatures (AETs) due to relatively large temperature changes in or flooding of the -36 foot elevation isothermal box during a LOCA/LOPA.

  17. Effects of Parallel Channel Interactions on Two-Phase Flow Split in ...

    African Journals Online (AJOL)

    The tests would aid the development of a realistic transient computer model for tracking the distribution of two-phase flows into the multiple parallel channels of a Nuclear Reactor, during Loss of Coolant Accidents (LOCA), and were performed at the General Electric Nuclear Energy Division Laboratory, California. The test ...

  18. Optical properties of nano-silicon

    Indian Academy of Sciences (India)

    Unknown

    distribution corresponding to strongly and weakly loca- lized or delocalized states in larger amorphous Si clusters. In hydrogenated clusters they show only the normal confine- ment effect and HOMO–LUMO gap values closer to the c-Si clusters. Similar results have also been obtained by. Allan et al (1997) for the average ...

  19. Temporal characteristics of aerosol physical properties at ...

    Indian Academy of Sciences (India)

    regions. It was reported that in coastal areas and inland seas the values of AOD are higher largely depending on the continental sources. In this con- text, the measurements at adjoining coastal loca- tions like Visakhapatnam during ICARB assume importance, particularly with reference to the cruise observations in the Bay ...

  20. Heat Transfer and Friction Characteristics of Artificially Roughened Duct used for Solar Air Heaters—a Review

    Science.gov (United States)

    Kumar, Khushmeet; Prajapati, D. R.; Samir, Sushant

    2018-02-01

    Solar air heater uses the energy coming from the sun to heat the air. The conversion rate of solar energy to heat depends upon the efficiency of the solar air heater and this efficiency can be increased by the use of artificial roughness on the surface of absorber plate. Various studies were carried out to analyse the effect of different roughness geometries on heat transfer and friction factor characteristics. The thermo-hydraulic performance of solar air heater can be evaluated in terms of effective efficiency, thermo-hydraulic performance parameter and exergetic efficiency. In this study various geometries used for artificial roughness and to improve the performance of solar air heaters were studied. Also correlations developed by various researchers are presented in this paper.

  1. Recommendations; Rekommendationer foer analys av spaenningsrespons i roersystem utsatta foer termohydrauliska transienter

    Energy Technology Data Exchange (ETDEWEB)

    Bjoerndahl, Olof; Letzter, Adam; Marcinkiewicz, Jerzy; Segle, Peter (Inspecta Nuclear AB, 104 25 Stockholm (Sweden))

    2007-03-15

    Transient thermohydraulic events often control the design of piping systems in nuclear power plants. Water hammers due to valve closure, pressure transients caused by steam collapse and pipe break all result in structural loads that are characterised by a high frequency content. What also characterises these pressures/forces is the specific spatial and time dependence that is acting on the piping system and found in the wave propagation in the contained fluid. The aim with this project has been to develop recommendations for analysis of the stress response in piping systems subjected to thermohydraulic transients. Basis for this work is that the so called two-step-method is applied and that the structural response is calculated with modal superposition. Derived analysis criteria are based on the assumption that the associated volume strain energy in the wave propagation for the contained fluid may be well defined by a parameter, here called epsilon{sub PN}. The stress response in the piping system is assumed to be completely determined with certain accuracy for that part of the volume strain energy in the wave propagation associated with this parameter. A comprehensive work has been done to determine the accuracy in loadings calculated with RELAP5. Properties such as period elongation and associated spurious oscillations in the pressure wave transient have been investigated. Furthermore, has the characteristics of the artificial numerical damping in RELAP5 been identified. Based on desired accuracy of the thermohydraulic analysis together with knowledge about the duration of the thermohydraulic perturbation, the lowest upper frequency limit f{sub Pipe}, in the modal base that is required for the structure model is calculated. With perturbation is meant such as a valve closure. According to suggested criteria and with the upper frequency limit set, the essential parameters i) largest size of the elements in the structure model and ii) the largest applicable time

  2. DUPIC fuel compatibility assessment; accident analysis of Wolsong-NPP for DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. H.; Kim, T. M.; Cho, C. H.; Hur, J. Y.; On, M. R.; Hwang, H. R.; Ahn, Z. K.; Kang, D. I. [KOPEC, Taejeon (Korea)

    2002-03-01

    Accident analysis of Wolsong NPP for DUPIC fuel is accomplished as a part of the nuclear fuel cycle technology development between the light water reactor and the heavy water reactor. Some analyses are performed for the thermohydraulic and radionuclide behaviour inside containment, radionuclide dispersion through atmosphere and public dose calculation after large loss of coolant accident. Wolsong 2 design data are used for containment model. For comparison with the result for natural uranium (NU) core, 100 % reactor outlet header break is selected for the limiting case which resulted in the significant public dose in Wolsong 2,3,4 FSAR. Single failure and dual failure cases, which are distinguished whether containment subsystem is working or not, are analyzed. PRESCON2 code for the thermohydraulic behaviour inside containment, SMART code for the radionuclide behaviour and PEAR code for atmospheric dispersion and the public dose calculation are used. 10 refs., 52 figs., 28 tabs. (Author)

  3. Scaled model studies of decay heat removal by natural convection for sodium cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. (Institut fuer Angewandte Thermo- und Fluiddynamik (IATF), Kernforschungszentrum Karlsruhe (Germany)); Weinberg, D. (Institut fuer Angewandte Thermo- und Fluiddynamik (IATF), Kernforschungszentrum Karlsruhe (Germany)); Marten, K. (Institut fuer Angewandte Thermo- und Fluiddynamik (IATF), Kernforschungszentrum Karlsruhe (Germany)); Schnetgoeke, G. (Institut fuer Angewandte Thermo- und Fluiddynamik (IATF), Kernforschungszentrum Karlsruhe (Germany))

    1993-06-01

    Thermohydraulic experiments were performed with water in order to simulate decay heat removal by natural convection in a pool-type sodium cooled reactor. Two water test rigs of different scales were used, namely, RAMONA (1:20) and NEPTUN (1:5). RAMONA was taken to study the transition from nominal operation by forced convection to decay heat removal operation by natural convection. Steady-state similarity tests were carried out in both facilities. All tests provide a basis for verification of computer programs. Calculations performed with the three-dimensional code FLUTAN proved that the thermohydraulic processes are quantitatively mastered, even for the very complex geometry of the NEPTUN test rig. (orig.)

  4. Aspects of brittle failure assessment for RPV

    Energy Technology Data Exchange (ETDEWEB)

    Zecha, H.; Hermann, T.; Hienstorfer, W. [TUeV SUeD Energietechnik GmbH Baden-Wuerttemberg, Filderstadt (Germany); Schuler, X. [Materialpruefungsanstalt, Univ. Stuttgart (Germany)

    2009-07-01

    This paper describes the process of pressurized thermal shock analysis (PTS) and brittle failure assessment for the reactor pressure vessel (RPV) of the nuclear power plants NECKAR I/II. The thermo-hydraulic part of the assessment provides the boundary conditions for the fracture mechanics analysis. In addition to the one dimensional thermo-hydraulic simulations CFD, analyses were carried out for selected transients. An extensive evaluation of material properties is necessary to provide the input data for a reliable fracture mechanics assessment. For the core weld and the flange weld it has shown that brittle crack initiation can be precluded for all considered load cases. For the cold and hot leg nozzle detailed linear-elastic and elasticplastic Finite Element Analyses (FEA) are performed to verify the integrity of the RPV. (orig.)

  5. Nuclear power plant cable materials :

    Energy Technology Data Exchange (ETDEWEB)

    Celina, Mathias C.; Gillen, Kenneth T; Lindgren, Eric Richard

    2013-05-01

    A selective literature review was conducted to assess whether currently available accelerated aging and original qualification data could be used to establish operational margins for the continued use of cable insulation and jacketing materials in nuclear power plant environments. The materials are subject to chemical and physical degradation under extended radiationthermal- oxidative conditions. Of particular interest were the circumstances under which existing aging data could be used to predict whether aged materials should pass loss of coolant accident (LOCA) performance requirements. Original LOCA qualification testing usually involved accelerated aging simulations of the 40-year expected ambient aging conditions followed by a LOCA simulation. The accelerated aging simulations were conducted under rapid accelerated aging conditions that did not account for many of the known limitations in accelerated polymer aging and therefore did not correctly simulate actual aging conditions. These highly accelerated aging conditions resulted in insulation materials with mostly inert aging processes as well as jacket materials where oxidative damage dropped quickly away from the air-exposed outside jacket surface. Therefore, for most LOCA performance predictions, testing appears to have relied upon heterogeneous aging behavior with oxidation often limited to the exterior of the cable cross-section a situation which is not comparable with the nearly homogenous oxidative aging that will occur over decades under low dose rate and low temperature plant conditions. The historical aging conditions are therefore insufficient to determine with reasonable confidence the remaining operational margins for these materials. This does not necessarily imply that the existing 40-year-old materials would fail if LOCA conditions occurred, but rather that unambiguous statements about the current aging state and anticipated LOCA performance cannot be provided based on

  6. Specialists' meeting on thermodynamics of FBR fuel subassemblies under nominal and non-nominal operating conditions. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    The purpose of the meeting was to provide a forum for exchange of information on thermo- and fluiddynamic investigations of LMFBR-subassembly. Special emphasis was placed on nominal and non-nominal conditions. The technical part of the meeting was divided into four sessions, as follows: status of the thermo- and fluiddynamic activities; physical and mathematical modelling of single phase; rod bundle thermohydraulics; experimental investigations; and future R and D. Separate abstracts are included for each of the papers.

  7. Physical metrology of aerosols; Metrologie physique des aerosols

    Energy Technology Data Exchange (ETDEWEB)

    Boulaud, D.; Vendel, J. [CEA Saclay, 91 - Gif-sur-Yvette (France). Inst. de Protection et de Surete Nucleaire

    1996-12-31

    The various detection and measuring methods for aerosols are presented, and their selection is related to aerosol characteristics (size range, concentration or mass range), thermo-hydraulic conditions (carrier fluid temperature, pressure and flow rate) and to the measuring system conditions (measuring frequency, data collection speed, cost...). Methods based on aerosol dynamic properties (inertial, diffusional and electrical methods) and aerosol optical properties (localized and integral methods) are described and their performances and applications are compared

  8. DABASCO Experiment Data Acquisition and Control System; Sistema de Toma de Datos y Control del Experimento DABASCO

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, J.; Artigao, A.; Barcala, J. M.; Oller, J. C. [Ciemat, Madrid (Spain)

    2000-07-01

    DABASCO experiment wants to study the thermohydraulic phenomena produced into the containment area for a severe accident in a nuclear power facility. This document describes the characteristics of the data acquisition and control system used in the experiment. The main elements of the system were a data acquisition board, PCI-MIO-16E-4, and an application written with LaB View. (Author) 5 refs.

  9. Scarabee programme and associated trials

    Energy Technology Data Exchange (ETDEWEB)

    Meyer-Heine, A.; Schmitt, A.P.; Aujollet, J.M.; Fortunato, M.; Chaudat, J.P. (Departement de Surete Nucleaire, CEA Fontenay-aux-Roses, 92 (France)); Penet, F. (CEA Cadarache, 13 - Saint-Paul-les-Durance (France)); Costa, J. (CEA Saclay, 91 - Gif-sur-Yvette (France)); Simeon, C. (SEPTEN-EDF, 92 - Clamart)

    This article describes the Scarabee set-up and the programme conducted in this experimental reactor, designed for the study of assembly accidents in the fast neutron field. Out-pile experiments carried out in support of the Scarabee programme are then presented. Two lines of research are distinguished: thermohydraulic (local obstructions, sodium boiling, melting of cladding) and fuel baths. Some studies more specific to molten fuels are also mentioned.

  10. Simulation of fluctuations of temperature and flow at the entrance of the nucleus of Trillo NPP; Simulacion de fluctuaciones de temperatura y caudal a la entrada del nucleo de C.N. Trillo

    Energy Technology Data Exchange (ETDEWEB)

    Bermejo, J. A.; Pelaez, S. B.; Lopez, A.; Ortego, A.

    2013-07-01

    One of the lines of work of the r and d program initiated in CNAT, to investigate the causes increasing the level of neutronic noise in C.N. Trillo, is the simulation of fluctuations thermohydraulics at the entrance of the neutron kernel with capability of simulation transient. The paper summarizes the status of this line of work and are presented some of the main results obtained so far.

  11. Simulation technology for training in the management of severe accidents in nuclear power; Tecnologia de simulacion para entrenamiento en gestion de accidentes severos en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Gil Moya, E.; Ruiz Martin, J. A.

    2012-07-01

    The objective of the project consists of the development of a module of severe accident based on the code Thermo-hydraulic MAAP and their integration in a Spanish CN training Simulator. Currently, stimulated the tools designed by Tecnatom aimed at training and assistance in the management of emergencies, complemented by the development of a dynamic interactive guides of severe accidents, thus constituting a set of aid for the operation.

  12. Physics of steam generators and visit of Saint-Marcel plant; La physique des generateurs de vapeur et la visite de l'Usine de Saint-Marcel

    Energy Technology Data Exchange (ETDEWEB)

    Gillet, N.; Gloaguen, C.; Holcblat, A. [FRAMATOME ANP, 92 - Paris La Defence (France); Borsoi, L. [CEA Saclay (SEMT/DYN), 91 - Gif sur yvette (France); Adobes, A.; David, F. [Electricite de France (EDF/RD), 75 - Paris (France); Greiner, E. [Electricite de France (EDF CIPN-CM), 13 - Marseille (France); Pascal-Ribot, S. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Gauchet, J.P. [Electricite de France (EDF/UTO/GVD), 93 - Noisy le Grand (France); Mercier, L. [Electricite de France (EDF/CAPE/GMC), 93 - Saint enis (France); Leomy, F. [FRAMATOME ANP, 71 - Chalon (France)

    2004-07-01

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in June 2004. The main topic was the physics of steam generators: 1 - description (G. Paudroux, J.Y. Guena, M. Petit); 2 - thermo-hydraulics (A. Holcblat, F. David, S. Pascal-Ribot); 3 - mechanics (N. Gillet, L. Borsoi, A. Adobes); 4 - monitoring and maintenance means (J.P. Gauchet, L. Mercier, F. Leomy); 5 - replacement (C. Gloaguen, E. Greiner). (J.S.)

  13. Correlation between PCT level CET and in the vessel of a reactor accidental sequences in stopping; Correlacion entre PCT, CET y nivel en la vasija de un reactor en secuencias accidentales de parada

    Energy Technology Data Exchange (ETDEWEB)

    Preciado, M.; Villanueva, J. F.; Carlos, S.; Martorell, S.

    2013-07-01

    In a setting of stop with discovery of kernel crash is of special interest the maximum temperature of fuel elements (PCT) pod. The difficulty to directly measure this temperature makes is look for the measurement of the temperature of exit of the nucleus (CET). This paper proposes to study the correlation between these basic parameters of measurement in a commercial plant, based on the results of the simulation of different cases through the Thermo-hydraulic TRACE code.

  14. Studies of the steam generator degraded tubes behavior on BRUTUS test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chedeau, C.; Rassineux, B. [EDF/DER/MTC, Moret Sur Loing (France); Flesch, B. [EDF/EPN/DMAINT, Paris (France)] [and others

    1997-04-01

    Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture.

  15. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  16. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  17. Morphological characteristics of glassy pyroclasts from the AD 1913 rift-edge phreatomagmatic eruption on Ambrym, Vanuatu

    Science.gov (United States)

    Nemeth, K.; Cronin, S. J.

    2009-12-01

    Thermohydraulic explosions are characterised by fine ash produced by brittle melt fragmentation and high kinetic energy release. Fine ash particles (peels appeared to form a mosaic-like surface (>4 Phi). A similar texture was not but micro-cracks been recognized on untreated coarse ash that was to enhanced by the ultrasonic bath treatment. The shape of these mosaic-like zones indicate brittle deformation, and their shape compares well with platy particles in the fine (>4 Phi) ash suggesting that they may have formed during transportation in the pyroclastic density current by spallation from the coarser particles. A gradual shift toward more brittle fragmentation shape characteristics has been recorded in the >2 Phi grain size fractions. This study gives evidence of the thermohydraulic magma/water interaction triggering fragmentation of the melt and distinguishes active, passive and abrasion-produced juvenile particles in individual beds. It seems that the proportion of these types of particles strongly depends on the physical conditions in the vent and the location of the deposit sampled relative to the vent. The presence of carbonate mud coatings and micro-mingling textures (peperitic micro-domain) in coarse ash may indicate the coarse mixing of the melt and coolant prior the thermohydraulic explosion.

  18. Experimental and analytical study on cesium iodide behavior in piping in wave experiment

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, A.; Igarashi, M.; Hashimoto, K.; Sugimoto, J. [Japan Atomic Energy Research Inst., Dep. of Reactor Safety Research, Tokai-mura (Japan); Yoshino, T. [Toshiba Advanced System Corp., Isago Kawasaki-ku (Japan)

    1996-12-01

    The WAVE (Wide range Aerosol model VErification) experiments have been performed at JAERI to investigate cesium iodide (CsI) deposition onto an inner surface of piping wall under typical severe accident conditions. The test facility consists of a dish containing CsI powder, electrical heaters and a straight pipe of 1.5 m in length with diameter of 0.042m. Nitrogen gas and superheated steam were injected into the pipe to carry the vaporized CsI and to simulate the thermohydraulic condition for the PWR hot-leg inlet. Analyses of the experiments have been conducted with a three-dimensional thermohydraulic code, SPRAC and the radionuclide behavior analysis codes, ART and VICTORIA. A clear difference was found in the deposition behavior between nitrogen and steam conditions as carrier gases. For nitrogen gas, the analyses well reproduced the experimental results by closely coupling the CsI behavior and the detailed thermohydraulic analyses. For steam carrier gas, on the contrary, the experimental results could not be well reproduced without the use of larger aerosol size. Since the observed enhancement of aerosol size in superheated steam cannot be explained by existing models, it is necessary to further investigate this mechanisms by experiment and analysis. (author) 34 figs., 23 refs.

  19. Assessing environmental effects on organic materials in cultural heritage

    DEFF Research Database (Denmark)

    Boyatzis, Stamatis; Ioakimoglou, Eleni; Facorellis, Yorgos

    2015-01-01

    Under the auspices of INVENVORG (Thales Research Funding Program – NRSF), and within a holistic approach for assessing environmental effects on organic materials in cultural heritage (CH) artefacts, the effect of artificial ageing on elemental and molecular damage and their effects on the structu......Under the auspices of INVENVORG (Thales Research Funding Program – NRSF), and within a holistic approach for assessing environmental effects on organic materials in cultural heritage (CH) artefacts, the effect of artificial ageing on elemental and molecular damage and their effects......, high performance liquid chromatography (HPLC) and Enzyme Linked Immunosorbent Assay (ELIZA) were realized. Results show damage within the inorganic and the organic matrix; incorporation of sulfur and nitrogen groups, minor reduction of specific aminoacids and changes in collagen integrity were...

  20. FAME Process: A Dedicated Development and V&V Process for FDIR

    Science.gov (United States)

    Guiotto, Andrea; De Ferluc, Regis; Bozzano, Marco; Cimatti, Alessandro; Gario, Marco; Yushtein, Yuri

    2014-08-01

    In the frame of the European Space Agency (ESA) studies, Thales Alenia Space Italia has carryed out a research - FAME - in collaboration with Fondazione Bruno Kessler and Thales Alenia Space France. The objective of the FAME project was to define a dedicated FDIR development, verification and validation process that can address the issues and shortcomings of the current industrial FDIR development practices. The ultimate goal was to allow for the consistent and timely FDIR conception, development, and Verification & Validation. A parallel objective of the study was the development of a toolset supporting the Process and enabling a coherent definition, specification, development, and V&V of the FDIR functionalities. It started in September 2013 and ended in May 2014.

  1. Democritos: preparing demonstrators for high power nuclear electric space propulsion

    OpenAIRE

    Masson, Frederic; Ruault, Jean-Marc; Worms, Jean-Claude; Detsis, Emmanouil; Beaurain, André; Lassoudiere, Francois; Gaia, Enrico; Tosi, Maria -Christina; Jansen, Frank; Bauer, Waldemar; Semenkin, Alexander; Tinsley, Tim; Hodgson, Zara

    2015-01-01

    The Democritos project aims at preparing demonstrators for a megawatt class nuclearelectric space propulsion. It is funded by Horizon 2020, the R&T program of the European Community. It is a new European and Russian project, including as partners: Nuclear National Laboratory (U.K.), DLR (Germany), The Keldysh Research Center (Russia), Thales Alenia Space Italia (Italy), Snecma (France), ESF (France) and CNES (France). IEAV (Brazil) will join as an observer. Democritos is the follo...

  2. The problem of the origin of philosophy in the text concerning history of philosophy in Turkish

    OpenAIRE

    Tufan Çötok

    2007-01-01

    Generally, the books concerning history of philosophy accept the proposition “the first principle of all things is water” that Thales said in 6th century BC as the birth of philosophical thought. What are the reasons and grounds of such an acceptance? Is to philosophise based on this acceptance a mental activity that extends from Ancient Greek to our times. Otherwise is there any possibility of philosophising differently in different cultures? All these questions are answered plai...

  3. DO-178C: A New Standard for Software Safety Certification

    Science.gov (United States)

    2010-04-26

    European Headquarters: 46 rue d’Amsterdam 75009 Paris France +33-1-4970-6716 (voice) +33-1-4970-0552 (FAX) www.adacore.com Ben Brosgol brosgol...Great Britain, France , Germany, Sweden • Some participation from Brazil and recently China Th i ki d f i ti t dree ma n n s o organ za ons represen e...Airframe industry and contractors (Boeing, Airbus, Lockheed, Rockwell Collins, Honeywell, Thales, Eurocopter , …) • Government agencies

  4. Thermal imaging for current D&S priorities

    Science.gov (United States)

    Craig, Robert; Parsons, John F.

    2012-11-01

    Supplying thermal imagers for today's operational needs requires flexibility, responsiveness and ever reducing costs. This paper will use the latest thermal imager development in the Catherine range from Thales UK to address the technical interactions with such issues as modularity, re-use, regions of deployment and supply chain management. All this is in the context of the increasingly public operations and the pressures on validating performance especially when weapon aiming is involved.

  5. Transgenic Arabidopsis Gene Expression System

    Science.gov (United States)

    Ferl, Robert; Paul, Anna-Lisa

    2009-01-01

    The Transgenic Arabidopsis Gene Expression System (TAGES) investigation is one in a pair of investigations that use the Advanced Biological Research System (ABRS) facility. TAGES uses Arabidopsis thaliana, thale cress, with sensor promoter-reporter gene constructs that render the plants as biomonitors (an organism used to determine the quality of the surrounding environment) of their environment using real-time nondestructive Green Fluorescent Protein (GFP) imagery and traditional postflight analyses.

  6. Case report

    African Journals Online (AJOL)

    abp

    14 janv. 2014 ... ont prouvé grâce à la technique de PCR l'association entre le. Campylobacter jéjuni et l'IPSID [5]. La toxine cytoléthale du C. jejuni engendre des cassures double-brin de l'ADN qui favorisent le développement de cellules B mutées, différenciées en cellules plasmocytaires aberrantes produisant des ...

  7. Detection of Buried Mines and Unexploded Ordnance (UXO)

    Science.gov (United States)

    2007-04-20

    maintenance and performance of the animals. q. Biological Systems: Plants Using a transgenic plant bioindicator implanted in the annual weed Thale...Provencal, Sze M. Tan, Eric R. Crosson, Alexander A. Kachanov, and Barbara A. Paldas. July 2003 Bananas , explosives and the future of cavity ring-down...development at the laboratory level. B-6 DATA SHEET B-4. BIOLOGICAL SYSTEMS: PLANTS Description of Technology Using a transgenic plant

  8. Anaximander of Miletus (c. 611-c. 547 BC)

    Science.gov (United States)

    Murdin, P.

    2000-11-01

    Greek philosopher, born in Miletus, pupil of THALES. Believed that the Earth was a cylinder with a diameter three times its height, unsupported, at the center of the universe. We live on the top end of the cylinder. His theory of the universe was that objects are formed from a vortex process by which light objects were flung out to their periphery. This separated hot and cold, moist and dry, and ...

  9. Proposal to negotiate the renewal of an existing contract for the repair, reconditioning or replacement of klystrons for CTF3

    CERN Document Server

    2006-01-01

    This document concerns the repair, reconditioning or replacement of pulsed klystrons being used in the CLIC Test Facility 3 (CTF3). The Finance Committee is invited to agree to the negotiation of the renewal of a contract for a period of five years with THALES (FR) for the repair, reconditioning or replacement of pulsed klystrons for an estimated total amount not exceeding 2 000 000 Swiss francs.

  10. Performance and Qualification of the Power Supply and Control Unit for the HEMP Thruster

    Science.gov (United States)

    Brag, R.; Herty, F.

    2014-08-01

    In 2013, Astrium GmbH delivered several flight model electronics for Electric Propulsion (EP) systems or corresponding components. One of the elements is a Power Supply and Control Unit (PSCU) for the Thales development "High Efficiency Multistage Plasma Thruster" (HEMP-T) (see Figure 1). This paper presents the PSCU specification and results of the qualification and acceptance phase of the EQM and the PFM.

  11. Kaitsevägi hangib iseliikuvaid suurtükke / Andres Einmann

    Index Scriptorium Estoniae

    Einmann, Andres

    2013-01-01

    Kaitseministeerium kulutab tuleval aastal investeeringuteks 43,26 miljonit eurot. Kavas on hankida uut sõjatehnikat: suurtükke ja lahingumasinaid, teine Thales Raytheoni Ground Master 403 tüüpi radar, Sisu XA-188 soomukeid. Suurim projekt on viie kasarmu ehitamine: aasta lõpuks valmivad neli kasarmut Võrus, Jõhvis ja Ämaris ning peale selle renoveeritakse üks hoone Miinisadamas

  12. BRIC-60: Biological Research in Canisters (BRIC)-60

    Science.gov (United States)

    Richards, Stephanie E. (Compiler); Levine, Howard G.; Romero, Vergel

    2016-01-01

    The Biological Research in Canisters (BRIC) is an anodized-aluminum cylinder used to provide passive stowage for investigations evaluating the effects of space flight on small organisms. Specimens flown in the BRIC 60 mm petri dish (BRIC-60) hardware include Lycoperscion esculentum (tomato), Arabidopsis thaliana (thale cress), Glycine max (soybean) seedlings, Physarum polycephalum (slime mold) cells, Pothetria dispar (gypsy moth) eggs and Ceratodon purpureus (moss).

  13. Muhus paigaldatakse õhuväe radarit

    Index Scriptorium Estoniae

    2012-01-01

    Esimene kahest õhuväe keskmaa-õhuseireradarist jõudis Prantsusmaalt Eestisse novembri lõpus, radar on transporditud Muhu saarele ning parasjagu käivad seadme paigaldustööd, ütles BNS-ile kaitseministeeriumi pressiesindaja. Paigaldamisele järgnevad häälestamine ja testimine. Tööd peaks Thales Ground Master 403 radar alustama veebruaris

  14. Treatment integrity of elaborated semantic feature analysis aphasia therapydelivered in individual and group settings

    OpenAIRE

    Kladouchou, V.; Papathanasiou, I.; Efstratiadou, E. A.; Christaki, V.; Hilari, K.

    2017-01-01

    Background & Aims\\ud \\ud This study ran within the framework of the Thales Aphasia Project that investigated the efficacy of elaborated semantic feature analysis (ESFA). We evaluated the treatment integrity (TI) of ESFA, i.e., the degree to which therapists implemented treatment as intended by the treatment protocol, in two different formats: individual and group therapy.\\ud \\ud Methods & Procedures\\ud \\ud Based on the ESFA manual, observation of therapy videos and TI literature, we developed...

  15. Development of the LPT9510 1 W Concentric Pulse Tube

    Science.gov (United States)

    Mullié, J. C.; Bruins, P. C.; Benschop, T.; Charles, I.; Coynel, A.; Duband, L.

    2006-04-01

    In order to provide cryogenic cooling for applications that are extremely sensitive to mechanical vibration, Thales Cryogenics has been delivering U-shape pulse tube cryocoolers since 2001. The disadvantage of the U-shape design is that the available regenerator volume is too limited if the application puts constrains on the overall diameter of the cold finger, thus limiting the coolers efficiency. As presented at CEC/ICMC 2003, Thales Cryogenics and CEA/SBT have achieved very good results with a large concentric pulse tube delivering 4W @ 77K driven by a flexure bearing compressor. Furthermore, the same team, together with Air Liquide DTA, developed a very efficient 1W pulse tube cooler for the ESA MPTC project. Based on the experiences obtained with those programs, Thales Cryogenics and CEA/SBT have now developed a small concentric pulse tube that is driven by a flexure bearing compressor. The result is a very compact and reliable cooler, with an efficiency that is nearly doubled compared to the U-shape version with the same overall external diameter dimensions. This paper describes the trade-offs that have been considered in the design phase, and gives a detailed overview of the test results, the status of the qualification program and a comparison with a comparable Stirling cold finger.

  16. How to manage MTTF larger than 30,000hr on rotary cryocoolers

    Science.gov (United States)

    Cauquil, Jean-Marc; Seguineau, Cédric; Martin, Jean-Yves; Van-Acker, Sébastien; Benschop, Tonny

    2017-05-01

    The cooled IR detectors are used in a wide range of applications. Most of the time, the cryocoolers are one of the components dimensioning the lifetime of the system. Indeed, Stirling coolers are mechanical systems where wear occurs on millimetric mechanisms. The exponential law classically used in electronics for Mean Time to Failure (MTTF) calculation cannot be directly used for mechanical devices. With new applications for thermal sensor like border surveillance, an increasing reliability has become mandatory for rotary cooler. The current needs are above several tens of thousands of continuous hour of cooling. Thales Cryogenics made specific development on that topic, for both linear and rotary applications. The time needed for validating changes in processes through suited experimental design is hardly affordable by following a robust and rigorous standard scientific approach. The targeted Mean Time to Failure (MTTF) led us to adopt an innovative approach to keep development phases in line with expected time to market. This innovative approach is today widespread on all of Thales Cryogenics rotary products and results in a proven increase of MTTF for RM2, RM3 and recently RM1. This paper will then focused on the current MTTF figures measured on RM1, RM2 and RM3. After explaining the limit of a conventional approach, the paper will then describe the current method. At last, the authors will explain how these principles are taken into account for the new SWaP rotary cooler of Thales Cryogénie SAS.

  17. Comparison of zircaloy-4 and zirlo by high-temperature oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, J. R.; Jeong, H.; Park, G. H. [Kyunghee Univ., Yongin (Korea, Republic of); Ryu, T. G. [FNC Technology, Seoul (Korea, Republic of)

    2003-10-01

    An experiment on the oxidation of Zirlo was performed in steam between 850 .deg. C and 1150 .deg. C. Zirlo cladding of the reactor fuel was used for the specimen. Temperature was the range appropriate for LOCA analyses of PWR. Tube-type Zirlo was manufactured as specimens by cutting, grinding, polishing and chemical polishing (etching). After the oxidation experiments, the mass increase of each specimen was measured, and the oxidation layer was observed by an optical microscope. Oxidation layer, and those of {alpha}-phase, {beta}-phase were also observed at the specimens oxidized at high temperatures. The hardness with respect to depth also was measured. It can be compared and verified safty of Zircaloy and Zirlo in LOCA according to the data of experiment.

  18. Analysis of ATLAS 6-inch cold leg break simulation with MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Se Yun; Jun, Hwang Yong; Ha, Sang Jun [Korea Electric Power Company, Daejeon (Korea, Republic of)

    2011-05-15

    A Domestic Standard Problem (DSP) exercise using ATLAS facility has been organized by KAERI. As the second DSP exercise, the 6-inch cold leg bottom break was determined. This experiment is the counterpart test to the DVI line break to verify the safety performance of the DVI method over the traditional CLI method. Compared with the large break LOCA, the phases of the small break LOCA prior to core recovery occur over a long period. The blowdown, natural circulation, loop seal clearance, boil-off, and core recovery phase should be investigated minutely with relevant models of safety analysis codes in order to predict these thermal hydraulic phenomena correctly. To investigate the ECC bypass phenomena, a finer study on the thermalhydraulic behavior in upper annulus downcomer was carried out

  19. LOFT reflood as a function of accumulator initial gas volume

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, H.F.

    1978-06-01

    The effect of the initial gas volume in the LOFT accumulators on the time to start of core reflood, after a LOCA, has been studied. The bases of the calculations are the data used and results presented in the Safety Analysis Report, Rev.1, August 1977, and the data in the RELAP and TOODEE2 program input and output listings. The results of this study show that an initial nitrogen volume of 12 cu ft, or more (at 600 psig initial pressure), would cause start of core reflood in time to prevent the cladding temperature from reaching 2200/sup 0/F. The 12 cu ft initial volume will expand from 600 psig, initial pressure, to about 10 psig (containment pressure shortly after start of LOCA is approximately 8 psig) when all ECC liquid has been expelled from the accumulator. This pressure margin is considered too small; the ECC flowrate will be zero before the accumulator is empty.

  20. Development of capability to Model A TRIGA reactor using ATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Davis, C.B.

    1985-05-01

    The capability to perform thermal-hydraulic analyses of a TRIGA reactor was demonstrated using the ATHENA computer code. TRIGA is an advanced reactor designed to produce electrical power while being inherently safe during reactivity accidents, loss-of-coolant accidents (LOCAs), and station blackout. The TRIGA system contains a water-filled primary system and a power conversion system that utilizes freon as the working fluid. An ATHENA model of a TRIGA-like reactor was developed. Calculations of a station blackout and a large-break LOCA were performed to demonstrate the capability of ATHENA to represent the TRIGA system. A mask of the TRIGA model and an interface with the Nuclear Plant Analyzer (NPA) were developed, allowing a graphic display of the calculated results on the NPA.

  1. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  2. Application of passive auto catalytic recombiner (PAR) for BWR plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, K.; Murano, K. [Tokyo Electric Power Co., Inc. (Japan); Yamanari, S. [Hitachi Cable, Ltd., Tokyo (Japan); Yamamoto, Y. [Toshiba Corp., Yokohama (Japan)

    2001-07-01

    The passive auto-catalytic recombiner (PAR), which can recombine flammable gases such as hydrogen and oxygen with each other to avoid an explosion in case of a loss-of-coolant accident (LOCA), installed in the primary containment vessel does not require a power supply or dynamic equipment, while the existing flammability gas control system (FCS) of most BWRs as an outer loop of the primary containment vessel needs them to make flammable gases circulate through blowers and heaters in the system. PAR offers a number of advantages over existing FCS, such as high reliability, low cost due to much smaller amount of materials needed, good maintainability, good operability in case of a LOCA, and smaller space for installation. An experimental study has been carried out for the purpose of solving the problems of applying PAR to Japanese BWR plants instead of existing FCS, in which we grasped the basic characteristics of PAR. (author)

  3. Useless Hearing in Male Emblemasoma auditrix (Diptera, Sarcophagidae) – A Case of Intralocus Sexual Conflict during Evolution of a Complex Sense Organ?

    OpenAIRE

    Reinhard Lakes-Harlan; Thomas Devries; Heiko Stölting; Andreas Stumpner

    2014-01-01

    Sensory modalities typically are important for both sexes, although sex-specific functional adaptations may occur frequently. This is true for hearing as well. Consequently, distinct behavioural functions were identified for the different insect hearing systems. Here we describe a first case, where a trait of an evolutionary novelty and a highly specialized hearing organ is adaptive in only one sex. The main function of hearing of the parasitoid fly Emblemasoma auditrix is to loca...

  4. Experimental observation of bias-dependent non-local Andreev reflection

    OpenAIRE

    Russo, S.; Kroug, M.; Klapwijk, T. M.; Morpurgo, A. F.

    2005-01-01

    We investigate transport through hybrid structures consisting of two normal metal leads connected via tunnel barriers to one common superconducting electrode. We find clear evidence for the occurrence of non-local Andreev reflection and elastic cotunneling through superconductor when the separation of the tunnel barrier is comparable to the superconducting coherence length. The probability of the two processes is energy dependent, with elastic cotunneling dominating at low energy and non-loca...

  5. ENGLISH TEACHERS’ PERCEPTIONS ABOUT THEIR TEACHING: USING ACTIVITY THEORY TO IDENTIFY CONTRADICTIONS

    OpenAIRE

    Ardi Marwan

    2009-01-01

    Abstract: This paper highlights the findings of a study which was undertaken at a vocational higher institution in Indonesia. The aim of the study was to explore English teachers’ perceptions about their English language teaching (ELT) in this institution. Activity theory (AT) was employed as the framework for guiding the study owing to the fact that its focus was on the identification of contradictions occurring in the activity system. From AT analysis, several contradictions could be loca...

  6. Image-Guided Cancer Nanomedicine

    OpenAIRE

    Dong-Hyun Kim

    2018-01-01

    Multifunctional nanoparticles with superior imaging properties and therapeutic effects have been extensively developed for the nanomedicine. However, tumor-intrinsic barriers and tumor heterogeneity have resulted in low in vivo therapeutic efficacy. The poor in vivo targeting efficiency in passive and active targeting of nano-therapeutics along with the toxicity of nanoparticles has been a major problem in nanomedicine. Recently, image-guided nanomedicine, which can deliver nanoparticles loca...

  7. Development Considerations of AREVA NP Inc.'s Realistic LBLOCA Analysis Methodology

    OpenAIRE

    Martin, Robert P.; Larry D. O'Dell

    2008-01-01

    The AREVA NP Inc. realistic large-break loss-of-coolant-accident (LOCA) analysis methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. This methodology conforms to the code scaling, applicability, and uncertainty (CSAU) methodology developed by the Technical Program Group for the United States Nuclear Regulatory Commission in the late 1980s. In addition, several practical consi...

  8. Runway Exit Designs for Capacity Improvement Demonstrations. Phase 2. Computer Model Development

    Science.gov (United States)

    1992-01-15

    simplify the number of internal computations of the model thus reducing CPU time. 2.3.2 Turnoff Algorithm Validation Procedure The validation of a...most inten- sive computational algorithm in REDIM 2.0, however, is the optimization of exit loca- tions consuming approximately 80% of the CPU time...with a 80286 or 80386 microprocessors *A math coprocessor *EGA or VGA color monitor -HP laser printer or Epson FX-80 dot-matrix printer -I MB minimum of

  9. Lower Oligocene bivalves of Ramanian Stage from Kachchh ...

    Indian Academy of Sciences (India)

    408. R P Kachhara et al. Ta b le. 1 . Check list a nd loca lity wise d istribution o f bivalves in. Kachchh. Lo cality. W a ior. B ermoti. Rakhdi. Sl. n o. S p ecies name. Lakhpat. W aior. Dam. S tream. D am. Geological range. 1. Chlamys. (. Chlamys. ) –. –. R. –. –. Low er. M io cene bhatiyaensis. Jain. 2. Chlamys. (. Aequ ipecten. ).

  10. The Role of Human Error in Design, Construction, and Reliability of Marine Structures.

    Science.gov (United States)

    1994-10-01

    Management Systems Assessment ( LACA ) (SAMSA) 287 Role of Human Error In Refiabihty of Marine Structures zCa zq ýWx zU<UW z lCA L--!-J zZ F. 0 OZ W Exq Ot 4-j...assessment module relative to each other, e.g., how GEFA, LOCA, VESA, LACA , OHFA, RIRA, LISA and SAMSA should be considered on a comparative basis. 5

  11. Myosin Va Plays a Role in Nitrergic Smooth Muscle Relaxation in Gastric Fundus and Corpora Cavernosa of Penis

    OpenAIRE

    Arun Chaudhury; Vivian Cristofaro; Carew, Josephine A.; Goyal, Raj K; Sullivan, Maryrose P.

    2014-01-01

    The intracellular motor protein myosin Va is involved in nitrergic neurotransmission possibly by trafficking of neuronal nitric oxide synthase (nNOS) within the nerve terminals. In this study, we examined the role of myosin Va in the stomach and penis, proto-typical smooth muscle organs in which nitric oxide (NO) mediated relaxation is critical for function. We used confocal microscopy and co-immunoprecipitation of tissue from the gastric fundus (GF) and penile corpus cavernosum (CCP) to loca...

  12. Impact of urbanization and gardening practices on common butterfly communities in France

    OpenAIRE

    Fontaine, Beno??t; Bergerot, Benjamin; Le Viol, Isabelle; Julliard, Romain

    2016-01-01

    Abstract We investigated the interacting impacts of urban landscape and gardening practices on the species richness and total abundance of communities of common butterfly communities across France, using data from a nationwide monitoring scheme. We show that urbanization has a strong negative impact on butterfly richness and abundance but that at a local scale, such impact could be mitigated by gardening practices favoring nectar offer. We found few interactions among these landscape and loca...

  13. Calibration and correction procedure for quantitative out-of-plane shearography

    OpenAIRE

    Zastavnik, Filip; Pyl, Lincy; Gu, Jun; Sol, Hugo; Kersemans, Mathias; van Paepegem, Wim

    2015-01-01

    Quantitative shearography applications continue to gain practical importance. However, a study of the errors inherent in shearography measurements, related to calibration of the instrument and correction of the results, is most often lacking. This paper proposes a calibration and correction procedure for the out-of-plane shearography with a Michelson interferometer. The calibration is based on the shearography measurement of known rigid-body rotations of a flat plate and accounts for the loca...

  14. First report of Moniliophthora roreri causing frosty pod rot (moniliasis disease) of cocoa in Mexico

    OpenAIRE

    Phillips Mora, W.; Couti{\\~n}o, A.; Coutiño, A.; Ortiz, Cf F.; López, Ap P.; J. HERNÁNDEZ; Aime, Mc C.

    2006-01-01

    Theobroma cacao, the source of cocoa or cacao, has been cultivated in Mexico for hundreds of years, with around 37,000 farms covering 62,000 Ha in Tabasco and Chiapas dedicated to its production. In March 2005, deformed and premature ripening cocoa pods were noted in the vicinity of Ignacio Zaragoza, Pichucalco in northern Chiapas. Chocolate-coloured lesions with creamy mycelium, darkening with age, mummies (shrivelled pods) and internal necrosis were also commonly observed. By April, loca...

  15. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  16. A Review of Global Learning & Observations to Benefit the Environment (GLOBE)

    Science.gov (United States)

    2010-04-01

    attributes that underpin inquiry-based learning. For example, science topics can either be globa %y or loca%y defined and data can be co%ected for scientific...developed during the planning stages and will continue as the campaign launches in September 2011. The research investigations will require students...support for planning, pilot testing, evaluating, and scaling up activities and events associated with the SCRC. The GPO follows UCAR business , human

  17. KARAKTERISASI RADIONUKLIDA PADA TIAP SUB-SISTEM KESELAMATAN REAKTOR DAYA BERBAHAN BAKAR MOX

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-04-01

    Full Text Available Pengganti Bahan bakar UO2, yang tergolong uranium pengkayaan rendah, adalah bahan bakar MOX yang mempunyai pengkayaan yang lebih tinggi. Bahan bakar MOX mempunyai kandungan plutonium dan nuklida dari golongan aktinida yang lebih tinggi dibandingkan bahan bakar UO2, yang akan menghasilkan karakteristik radionuklida yang berbeda untuk setiap sub-sistem reaktor daya. Analisis radionuklida untuk setiap sub-sistem keselamatan pada reaktor daya berbahan bakar MOX dilakukan untuk mengetahui karakteristik radionuklida khususnya plutonium dan aktinida yang akan menimbulkan dampak radiasi dari lepasan radionuklida tersebut. Analisis dilakukan dengan cara menghitung dan mengamati radionuklida untuk setiap sub-sistem keselamatan pada operasi normal dan kecelakaan (small LOCA, large LOCA, severe accident untuk reaktor PWR berkapasitas 1000 MWe. Disimpulkan bahwa penggunaan bahan bakar MOX dapat menambah konsekuensi radiologis ke lingkungan dan masyarakat, terutama karena inventori yang lebih besar termasuk dari radionuklida transuranic dan dari golongan aktinida, antara lain: Pu-239, Am-241, Cm-242, Pu-240, Pu-241 dan Pu-242. Kata kunci: karakteristik nuklida, reaktor daya, bahan bakar, MOX   Substitute UO2 fuel that low enrichment of uranium is that MOX fuel has a higher enrichment. MOX fuel has a content of plutonium and actinide nuclides a higher than UO2 fuel, which will produce different characteristics of radionuclides for each sub-system of power reactors. Analyzis of radionuclide for each safety sub-system at MOX power reactor aims to determine the characteristics of radionuclides, especially plutonium and actinides consequences. Analyzis has done by calculating and observing the radionuclide for each safety sub-system in normal operation and accident (small LOCA, large LOCA, and severe accident on PWR-1000 reactors. It can concluded that the use of MOX fuel can add to the radiological consequences to the environment and public, mainly because a

  18. Profiling of Amatoxins and Phallotoxins in the Genus Lepiota by Liquid Chromatography Combined with UV Absorbance and Mass Spectrometry

    OpenAIRE

    R. Michael Sgambelluri; Sara Epis; Davide Sassera; Hong Luo; Angelos, Evan R; Walton, Jonathan D.

    2014-01-01

    Species in the mushroom genus Lepiota can cause fatal mushroom poisonings due to their content of amatoxins such as α-amanitin. Previous studies of the toxin composition of poisonous Lepiota species relied on analytical methods of low sensitivity or resolution. Using liquid chromatography coupled to UV absorbance and mass spectrometry, we analyzed the spectrum of peptide toxins present in six Italian species of Lepiota, including multiple samples of three of them collected in different loca...

  19. "Con notable daño del buen servicio": sobre la locura femenina en la primera mitad del siglo XX en Bogotá

    Directory of Open Access Journals (Sweden)

    María Angélica Ospina Martínez

    2006-06-01

    Full Text Available This essay proposes some analytical possibilities concerning the relationship between madness and gender in the history of mental asylums in the mid-20th century in Colombia. The research is based upon institutional documents and medical reports of the Asilo de Locas de Bogotá. It outlines the relation between social conceptions of female madness and the application of specific diagnosis and treatments in 1930s and 1940s.

  20. Estimates of Surface Drifter Trajectories in the Equatorial Atlantic: A Multi-model Ensemble Approach

    Science.gov (United States)

    2012-01-01

    trajectory end point that contained 80% of the actual drifter loca- tions. We could have shown PDFs, but we preferred to present the results with...eddy field of SURCOUF was actually inferior to the OGCMs, but the deficiency was compensated for by the superior mean field? We confirmed that this...Barron C, Carncs M, Lee C (2002) The modular ocean data assimilation system ( MODAS ). J Atmos Ocean Technol 19:240-252 Gentemann C Minnett P

  1. Population distribution in coastal of Linhares-ES

    OpenAIRE

    Felipe Pinto Gonçalves

    2014-01-01

    The article analyzes the distribution of population in the coastal districts of the municipality of Linhares-ES. At first, the coastal quota is contextualized in relation to the total population. Then, a retrospective of the peopling of the districts is presented. Finally, from the analysis of orthophotos, field observations and statistical data, are characterized forms of organization of housing and areas of population distribution. It is observed that most of the population is loca...

  2. Egg distribution, bottom topography and small-scale cod population structure in a coastal marine system

    OpenAIRE

    Knutsen, Halvor; Olsen, Espen Moland; Ciannelli, Lorenzo; Espeland, Sigurd Heiberg; Knutsen, Jan Atle; Simonsen, Jan Henrik; Skreslet, Stig; Stenseth, Nils Christian

    2007-01-01

    Coastal marine species with pelagic egg and larval stages, such as the Atlantic cod Gadus morhua, can be structured into genetically distinct local populations on a surprisingly small geographic scale considering their dispersal potential. Mechanisms responsible for such small-scale genetic structure may involve homing of adults to their natal spawning grounds, but also local retention of pelagic eggs and larvae. For example, spawning within sheltered fjord habitats is expected to favour loca...

  3. Magnetic Resonance Imaging Verification of a Case of Sacrococcygeal Teratoma

    OpenAIRE

    Dedushi, Kreshnike; Kabashi, Serbeze; Mucaj, Sefedin; Ramadani, Naser; Hoxhaj, Astrit; Shatri, Jeton; Hasbahta, Gazmend

    2016-01-01

    Although rare, sacrococcygeal teratoma is the most common congenital neoplasm, occurring in 1 in 40,000 infants. Approximately 75% of affected infants are female. The aim of the present study was to correlate ultrasonography and magnetic resonance imaging (MRI) findings in patients with fetal sacrococcygeal teratoma. Three pregnant women in 27th week of gestation underwent fetal MRI after ultrasonography examination, with findings suggestive for fetal sacrococcygeal teratoma. Tumor size, loca...

  4. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E.; Ivanova, A. [Energoproekt, Sofia (Bulgaria)

    1995-12-31

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment.

  5. Untitled

    African Journals Online (AJOL)

    les services d'Anatomie pathologique de. Côte d'lvoire. Les renseignements suivants ont été évalués: l'âge, l'origine géographi- que des patients ainsi que les caractéristi- ques cliniques des tumeurs telles leur loca- lisation, leur nature et leur type histologi- que. Certaines données telles que les mar- queurs tumoraux et la ...

  6. Asthma: Eosinophil Disease, Mast Cell Disease, or Both?

    OpenAIRE

    Bradding, Peter

    2008-01-01

    Although there is much circumstantial evidence implicating eosinophils as major orchestrators in the pathophysiology of asthma, recent studies have cast doubt on their importance. Not only does anti-interleukin-5 treatment not alter the course of the disease, but some patients with asthma do not have eosinophils in their airways, whereas patients with eosinophilic bronchitis exhibit a florid tissue eosinophilia but do not have asthma. In contrast, mast cells are found in all airways and loca...

  7. American cities, global networks: mapping the multiple geographies of globalization in the Americas

    OpenAIRE

    Toly, N.J.; Bouteligier, S.; Smith, G.; Gibson, B.

    2012-01-01

    The mapping of advanced producer and financial service firms across global cities began to increase understanding of the role of cities in global governance, the presence and influence of cities in the shifting architecture of global political economy, and the role of globalization in shaping the landscape of local and re- gional governance. The literature that emerged from such studies has also emphasized 1) increasing levels of inequality in global cities and 2) attendant contests over loca...

  8. A Review of the Historical Development and Contemporary Management Structure of Military Club Systems.

    Science.gov (United States)

    1980-06-01

    process, are customarily briefed by their counterparts as to the loca- I tions of the local military retail merchandise outlets (ex- changes), military... supermarkets (commissaries), military recreational facilities, and military clubs. These activities, which are lumped together under the generic term...The sutler system was replete with many abuses- -notably highI:prices, shoddy merchandise , and usurous interest rates. Num- erous cases of fraud and

  9. Lightning in the Ionosphere with C/NOFS

    Science.gov (United States)

    2012-08-25

    loca lightning, electric field, ionosphere , ionospheric plasma, plasma irregularities U U U U Prof. Robert H. Holzworth 206 685 7410 Reset INSTRUCTIONS... ionospheric plasma density irregularities and the occurrence of optical lightning activity; and the observation that lightning electric fields were often the...Geophysical Union, Fall Meeting 2011, abstract #AE21A-0239 On the Relationship between Lightning and Equatorial Ionosphere Density Irregularities , McCarthy

  10. Latex Imaging by Environmental STEM: Application to the Study of the Surfactant Outcome in Hybrid Alkyd/Acrylate Systems

    OpenAIRE

    Faucheu, Jenny; Chazeau, Laurent; Gauthier, Catherine; Cavaille, Jean-Yves; Goikoetxea, Monika; Minari, Roque; Asua, Jose M.

    2009-01-01

    International audience; Among other uses. latexes are a successful alternative to solvent-borne binders for coatings. Efforts are made to produce hybrid nanostructured latexes containing an acrylic phase and an alkyd phase, However, after the film-forming process, the surfactant used to stabilize these latexes remains in the film, and its location can have a drastic effect on the application properties. Among the processing parameters, the alkyd hydrophobicity can strongly influence this loca...

  11. Proceedings of Workshop 15 of the COSPAR Meetings Held in Toulouse, France on 30 June-12 July 1986. Chapter 2. Reference Atmospheres and Thermospheric Mapping,

    Science.gov (United States)

    1988-01-21

    muLiitudt of the 6r4adients in uuL..-er isi~et~ in tL u off sctt-llitu dacta thaz in the cozt; of rucktt; onda cLta.Thd uiaini- 1:." J-. WJ.tur i.s loca±ted...zonal wind amplitudes plotted as a funcion of latitude; it is evident that the maximum response is in the meridional component at southern low-to-mid

  12. Design of an off-grid hybrid PV/wind power system for remote mobile base station: A case study

    OpenAIRE

    Mulualem T. Yeshalem; Baseem Khan

    2017-01-01

    There is a clear challenge to provide reliable cellular mobile service at remote locations where a reliable power supply is not available. So, the existing Mobile towers or Base Transceiver Station (BTSs) uses a conventional diesel generator with backup battery banks. This paper presents the solution to utilizing a hybrid of photovoltaic (PV) solar and wind power system with a backup battery bank to provide feasibility and reliable electric power for a specific remote mobile base station loca...

  13. Implementasi Model Pembelajaran Bahasa Inggris di SD di Surakarta

    OpenAIRE

    Kaltsum, Honest Ummi; Utami, Ratnasari Diah

    2015-01-01

    This research aims to investigate the English learning model applied in elementary school in Surakarta. The population of this research is elementary schools in Surakarta and four elementary schools act as a sample. This research applies qualitative approaches using interview, observation, and documentation to obtain the data. The findings show that elementary schools in Surakarta apply one of the KTSP Curriculum and 2013 Curriculum. Those who apply 2013 Curriculum, consider English as a loca...

  14. Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-08-29

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U3Si2 at LWR conditions. The fission gas behavior of U3Si2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U3Si2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U3Si2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U3Si2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

  15. MALE HOMOSEXUAL IDENTITIES, RELATIONSHIPS, AND PRACTICES AMONG YOUNG MEN WHO HAVE SEX WITH MEN IN VIETNAM: IMPLICATIONS FOR HIV PREVENTION

    OpenAIRE

    Ngo, Duc Anh; Ross, Michael W; Phan, Ha; Ratliff, Eric A.; Trinh, Thang; Sherburne, Lisa

    2009-01-01

    Rapid socioeconomic transformation in Vietnam in last 15 years has been followed by more liberation of sexual expression and representation of sexual identity among young people. There has been an increase in the visibility of homosexual men in major cities of Vietnam who were largely an unknown population until the emergence of the HIV epidemic. Men who have sex with men (MSM) are now considered as one of the target groups in many HIV prevention programs. This qualitative study examines loca...

  16. Mise au point

    African Journals Online (AJOL)

    7 mai 2012 ... EPIdEMIOlOgIE ET ETIOPAThOgENIE. Les angiodysplasies osseuses des maxillaires sont rares. Les premiers cas rapportés sont ceux de Berard qui a décrit en 1842 une localisation au niveau du maxillaire supérieur et de Stanley qui a rapporté en 1849 une loca- lisation mandibulaire. une prédisposition ...

  17. Post-translational Modifications Regulate Class IIa Histone Deacetylase (HDAC) Function in Health and Disease*

    OpenAIRE

    Mathias, Rommel A.; Amanda J Guise; Cristea, Ileana M.

    2015-01-01

    Class IIa histone deacetylases (HDACs4, -5, -7, and -9) modulate the physiology of the human cardiovascular, musculoskeletal, nervous, and immune systems. The regulatory capacity of this family of enzymes stems from their ability to shuttle between nuclear and cytoplasmic compartments in response to signal-driven post-translational modification. Here, we review the current knowledge of modifications that control spatial and temporal histone deacetylase functions by regulating subcellular loca...

  18. Simulation of a SGTR severe PWR-W with the MELCOR code; Simulacion de un SGTR severo en un PWR-W con el codigo MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, A. J.; Israelsson, C.; Jimenez, G.

    2013-07-01

    The type SGTR accident is a case of loss of coolant accident small features which make it necessary to differentiate and evolution of classical studies LOCA sequence type. To simulate this type of accident has chosen the MELCOR code, which aims to study the progression of severe accidents in LWR plants. It has been developed by Sandia National Laboratories for the United States Nuclear Regulatory Commission.

  19. Elaine Frantz Parsons, Ku-Klux: The Birth of the Klan during Reconstruction

    OpenAIRE

    Grandi, Elisa

    2017-01-01

    Il libro di Elaine Frantz Parsons, professoressa associata della Duquesne University (Pittsburgh, Pennsylvania) fornisce il primo resoconto dettagliato delle origini e dell’affermazione del Ku Klux Klan all’indomani della Guerra Civile. A differenza di altri episodi di violenza razziale che caratterizzarono gli anni immediatamente successivi alla sconfitta degli Stati del Sud, il Klan si affermò come un fenomeno radicalmente diverso, poggiando allo stesso tempo su di un forte radicamento loca...

  20. Reconstruction of Acoustic Exposure on Orcas in Haro Strait

    Science.gov (United States)

    2009-01-01

    Resident killer whales (Orcinus orca) (J pod).1 The class shadowed the J pod from their boat , recording its behavior, the GPS loca- tion of the... boat , and the time of day. Figure 1 shows the tracks of USS Shoup and Dr. Bain’s boat shadowing the J pod overlaid on the bathymetry. Additionally...reverberation, i.e., sound energy that scattered from interactions with the ocean surface and bottom. The time series showed up to 19 s of

  1. People, states and hope : cosmopolitanism and the future of international law

    OpenAIRE

    Redmond, Trevor

    2010-01-01

    The term “cosmopolitanism” has been said to have become a key word of our time. The thesis presented here is that the political philosophy of cosmopolitanism is of relevance to, and has a history within, international law, such that it offers international law some hope of moving beyond a concern solely to secure the formal principles of external liberty between states, towards a greater concern for establishing a minimum level of material welfare for all individuals, regardless of their loca...

  2. Economic Evaluation of Project Site Using Cardinal Numbers Approach

    OpenAIRE

    Gul, Ejaz

    2013-01-01

    Selection of suitable site for construction project is essential since it has strong linkage with service life of the project. Recent fast developments in construction technology consider only the technical suitability of the project site but ignore the economic suitability. There can be many instances when a site may be suitable from technical point but not from economic point of view and vice versa. This research is about finding economic suitability of three different project sites loca...

  3. Sensitive and specific detection of Xanthomonas campestris pv. pelargonii with DNA primers and probes identified by random amplified polymorphic DNA analysis.

    OpenAIRE

    Manulis, S; Valinsky, L; Lichter, A; Gabriel, D. W.

    1994-01-01

    The random amplified polymorphic DNA method was used to distinguish strains of Xanthomonas campestris pv. pelargonii from 21 other Xanthomonas species and/or pathovars. Among the 42 arbitrarily chosen primers evaluated, 3 were found to reveal diagnostic polymorphisms when purified DNAs from compared strains were amplified by the PCR. The three primers revealed DNA amplification patterns which were conserved among all 53 strains tested of X. campestris pv. pelargonii isolated from various loca...

  4. Fatigue Lifetime Estimation Based on Rainflow Counted Data Using the Local Strain Approach

    OpenAIRE

    Dreßler, Klaus; Hack, Michael

    1995-01-01

    In the automotive industry both the loca l strain approach and rainflow counting are well known and approved tools in the numerical estimation of the lifetime of a new developed part especially in the automotive industry. This paper is devoted to the combination of both tools and a new algorithm is given that takes advantage of the inner structure of the most used damage parameters.

  5. Sensitivity Analysis of RCW Temperature on the Moderator Subcooling Margin for the LBLOCA of Wolsong NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Si Won; Kim, Jong Hyun; Choi, Sung Soo [Atomic Creative Technology Co., Daejeon (Korea, Republic of); Kim, Sung Min [Central Research Institute, Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-05-15

    Moderator subcooling margin has been analyzed using the MODTURC{sub C}LAS code in the Large LOCA FSAR PARTs C and F. Performance of moderator heat exchangers depends on RCW (Raw reCirculated Water) temperature. And also the temperature is affected by sea water temperature. Unfortunately, sea water temperature is gradually increasing by global warming. So it will cause increase of RCW temperature inevitably. There is no assessment result of moderator subcooling with increasing RCW temperature even if it is important problem. Therefore, sensitivity analysis is performed to give information about the relation between RCW temperature and moderator subcooling in the present study. The moderator subcooling margin has to be ensured to establish the moderator heat removal when Large LOCA with LOECI and Loss of Class IV Power occurs. However, sea water temperature is increasing gradually due to global warming. So it is necessary that sensitivity analysis of RCW temperature on the moderator subcooling margin to estimate the availability of the moderator heat removal. In the present paper, the moderator subcooling analysis is performed using the same methodology and assumptions except for RCW temperature used in FSAR Large LOCA PART F.

  6. Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swelling and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.

  7. Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

    Full Text Available The loss-of-coolant accident (LOCA simulation in the boiling water reactor (BWR of Laguna Verde Nuclear Power Plant (LVNPP at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

  8. Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yuming, E-mail: Yuming.chen@kit.edu; Ghidersa, Bradut-Eugen; Jin, Xue Zhou

    2016-11-01

    Highlights: • This paper presents transient CFD analyses on the cooling channels of the DEMO HCPB FW for accidental scenarios LOCA and LOFA. • In both LOCA & LOFA, the wall temperature increases quickly to an unacceptable level within seconds. • If the coolant flow rate is maintained at a half of nominal value in case of LOFA (partial LOFA), the wall temperature rises much slower, but will still leads to a damage of structure within minutes. • The simulated heat transfer coefficients were compared with empirical correlations. - Abstract: Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

  9. Test facility for rewetting experiments at CDTN

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C., E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2015-07-01

    One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

  10. Analysis of small break loss of coolant accident for Chinese CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

    2016-05-15

    This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

  11. FLECHT SEASET program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hochreiter, L E

    1985-11-01

    This report presents the highlights and main findings of the USNRC, EPRI, and Westinghouse cooperative FLECHT SEASET program. The report indicates areas in which the results of the program can contribute to revising the current licensing requirements for Loss of Coolant (LOCA) safety analysis for PWRs. Also identified are several technical areas in which the new FLECHT SEASET data and analysis can lead to improved safety analysis modeling, and thereby to predicted PWR response for postulated accident scenarios. Significant progress has been made in the modeling areas of nonequilibrium dispersed two-phase flow during reflood. Improved models and understanding of this rod bundle cooling regime are summarized in this report. Another important result of the FLECHT SEASET program arises from the natural circulation test series, which investigated single-phase, two-phase, and reflux condensation cooling modes of a scaled PWR under small-break LOCA conditions. The tests and subsequent analysis constitute one of few complete sets of data for these cooling modes in which full-height, multitube steam generators with sufficient instrumentation were used to examine primary-to-secondary heat transfer in the generators. It is believed that the natural circulation test data will be extremely useful to benchmark the improved post-TMI small-break LOCA computer codes. 170 figs., 13 tabs.

  12. Validation of the fast-running in-vessel model ASTRID for predicting the radioactive releases to the containment

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko, M. [VTT Processes (Finland); Schmuck, P. [Forschungszentrum Karlsruhe, Karlsruhe (Germany)

    2004-07-01

    The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) process model is used for the faster than real-time prediction of the radioactivity released into the containment and further into the environment in case of an emergency situation in a light water reactor. Combined together with the containment module COCOSYS the model can predict the entire radioactivity release chain from the primary system to the containment and further into the environment. In the paper the ASTRID thermohydraulic module PROCESS is presented shortly. The thermohydraulic part is a fast running solution for the drift-flux based thermohydraulics. In high temperatures the core degradation leading to the melt pool formation in the reactor barrel and reactor vessel lower head is calculated in the in-vessel module RELOMEL. Finally after the reactor vessel wall has been eroded due to the molten corium in the lower plenum, the massive radioactivity release occurs into the containment. But even before this scenario the radioactivity may be transported from the superheated core to the containment by the coolant. The reference plants for the development have been the Westinghouse type 4-loop PWR, the French type 3-loop PWR, The German type 4-loop Konvoi PWR, the Loviisa VVER type PWR, and the Olkiluoto type internal pump BWR. The reference code for the DBA thermal hydraulics has been the SMABRE code. In the developmental assessment the capability of the rough nodalization of ASTRID has been tested against the SMABRE nodalization describing the plants with 50 - 500 nodes. For the developmental assessment of the in-vessel severe accident the sample cases are calculated with MELCOR. The more thorough validation is based on the internationally known system codes, RELAP5, MELCOR, CATHARE and ATHLET. In the validation the most problematic area is the radioactivity transport into the containment. This part of the validation is done with the integrated code system. (authors)

  13. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    Science.gov (United States)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  14. SPIDER beam dump as diagnostic of the particle beam

    Energy Technology Data Exchange (ETDEWEB)

    Zaupa, M., E-mail: matteo.zaupa@igi.cnr.it; Sartori, E. [Università degli Studi di Padova, Via 8 Febbraio 2, Padova 35122 (Italy); Consorzio RFX, Corso Stati Uniti 4, Padova 35127 (Italy); Dalla Palma, M.; Brombin, M.; Pasqualotto, R. [Consorzio RFX, Corso Stati Uniti 4, Padova 35127 (Italy)

    2016-11-15

    The beam power produced by the negative ion source for the production of ion of deuterium extracted from RF plasma is mainly absorbed by the beam dump component which has been designed also for measuring the temperatures on the dumping panels for beam diagnostics. A finite element code has been developed to characterize, by thermo-hydraulic analysis, the sensitivity of the beam dump to the different beam parameters. The results prove the capability of diagnosing the beam divergence and the horizontal misalignment, while the entity of the halo fraction appears hardly detectable without considering the other foreseen diagnostics like tomography and beam emission spectroscopy.

  15. Nonlinear dynamo action in a precessing cylindrical container.

    Science.gov (United States)

    Nore, C; Léorat, J; Guermond, J-L; Luddens, F

    2011-07-01

    It is numerically demonstrated by means of a magnetohydrodynamics code that precession can trigger the dynamo effect in a cylindrical container. When the Reynolds number, based on the radius of the cylinder and its angular velocity, increases, the flow, which is initially centrosymmetric, loses its stability and bifurcates to a quasiperiodic motion. This unsteady and asymmetric flow is shown to be capable of sustaining dynamo action in the linear and nonlinear regimes. The magnetic field thus generated is unsteady and quadrupolar. These numerical evidences of dynamo action in a precessing cylindrical container may be useful for an experiment now planned at the Dresden sodium facility for dynamo and thermohydraulic studies in Germany.

  16. Nuclear cogeneration based on HTR technology

    Energy Technology Data Exchange (ETDEWEB)

    Haverkate, B.R.W.; Van Heek, A.I.; Jehee, J.N.T

    1998-03-01

    The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of an Inherently safe Nuclear COGENeration plant (INCOGEN). Technical, economical and licensing issues have been investigated of the INCOGEN design which comprises pebble fuel cooled by helium, and directly coupled with a helium compressor and turbine. Thermohydraulic and reactor physics analyses show favourable control characteristics during normal operation and a benign response to loss of helium coolant and loss of flow conditions. Throughout the response on these highly infrequent conditions, ample margin exists between the highest fuel temperatures and the temperature above which fuel degradation will occur. 7 refs.

  17. Generation IV: new reactor systems; Neue Reaktorsysteme innerhalb der Generation IV Initiative

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). IKET; Hofmeister, J. [RWE Power AG, Regenerative Stromerzeugung, Essen (Germany); Tromm, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Programm Nukleare Sicherheitsforschung

    2006-07-01

    Generation IV, an initiative for international cooperation in nuclear technology, was launched by 10 states in 2000 and joined by Euratom in July 2003. Its aim is to assess nuclear energy systems complying with future safety, disposal, proliferation, and public acceptance requirements. The Forschungszentrum Karlsruhe focuses on design, thermohydraulics, and neutron kinetics. Work is mainly devoted to the high-performance light water reactor (HPLWR) with supercritical steam conditions. Thus, competence can be maintained, as the HPLWR issues qualify for later work in nuclear industry. (orig.)

  18. Supercritical Helium Cooling of the LHC Beam Screens

    CERN Document Server

    Hatchadourian, E; Tavian, L

    1998-01-01

    The cold mass of the LHC superconducting magnets, operating in pressurised superfluid helium at 1.9 K, must be shielded from the dynamic heat loads induced by the circulating particle beams, by means of beam screens maintained at higher temperature. The beam screens are cooled between 5 and 20 K by forced flow of weakly supercritical helium, a solution which avoids two-phase flow in the long, narr ow cooling channels, but still presents a potential risk of thermohydraulic instabilities. This problem has been studied by theoretical modelling and experiments performed on a full-scale dedicated te st loop.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  20. Analyzing different HPCI operation modes simulated with ATHLET-CD regarding possible core degradation phenomena in Fukushima-Daiichi unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Bratfisch, Christoph; Koch, Marco K. [Ruhr-Univ. Bochum (Germany). Reactor Simulation and Safety Group

    2017-02-15

    For extented application and analyses of the severe accident code ATHLET-CD, the course of the invessel accident in Unit 3 of Fukushima-Daiichi is simulated in the frame of the research project SUBA as a part of the BMBF sponsored collaborative project WASA-BOSS (Weiterentwicklung und Anwendung von Severe Accident Codes - Bewertung und Optimierung von Stoerfallmassnahmen). Investigations, carried out by TEPCO, had shown that the High-Pressure Coolant Injection system (HPCI) might have stopped earlier than expected. A parameter variation was performed to analyze the impact of the tripped HPCI injection regarding the thermohydraulic behaviour as well as the core degradation phenomena.

  1. Numerical Analysis on the Calandria Tubes in the Moderator of a Heavy Water Reactor Using OpenFOAM and Other Codes

    Science.gov (United States)

    Chang, Se-Myong; Kim, Hyoung Tae

    2014-06-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors.

  2. The equation of state of liquid Flibe

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiang M.; Schrock, V.E.; Peterson, P.F.

    1991-08-16

    Flibe (Li{sub 2}BeF{sub 4}) is a candidate material for the liquid blanket in the HYLIFE-2 fusion reactor. The thermodynamic properties of the material are important for the study of thermohydraulic behavior of the concept design, including the compressible analysis of the blanket isochoric heating problem and resulting jet breakup. The equation of state provides the relationship between all the thermodynamic properties. Previously, a soft sphere model of liquid equation of state was used for describing a number of liquid metals. In this paper we have fitted the available experimental data for liquid Flibe with a modified soft sphere model. 5 refs.

  3. Cryogenic generator cooling

    Science.gov (United States)

    Eckels, P. W.; Fagan, T. J.; Parker, J. H., Jr.; Long, L. J.; Shestak, E. J.; Calfo, R. M.; Hannon, W. F.; Brown, D. B.; Barkell, J. W.; Patterson, A.

    The concept for a hydrogen cooled aluminum cryogenic generator was presented by Schlicher and Oberly in 1985. Following their lead, this paper describes the thermal design of a high voltage dc, multimegawatt generator of high power density. The rotor and stator are cooled by saturated liquid and supercritical hydrogen, respectively. The brushless exciter on the same shaft is also cooled by liquid hydrogen. Component development testing is well under way and some of the test results concerning the thermohydraulic performance of the conductors are reported. The aluminum cryogenic generator's characteristics are attractive for hydrogen economy applications.

  4. Preliminary evaluation of a pre-industrial air-cooled LiBr-H2O small capacity absorption machine

    OpenAIRE

    Farnós Baulenas, Joan; Castro González, Jesús; Morales Ruíz, Sergio; García Rivera, Eduardo; Oliva Llena, Asensio

    2014-01-01

    The paper describes the thermal design, and evaluates the preliminary operational results of a small capacity pre-industrial LiBr-H2O air-cooled absorption machine in order to validate a numerical model and apply it to simulate a new 7kW LiBr-H2O air-cooled absorption chiller, conceived for low temperature driven, solar cooling or wasted heat. This numerical model is able to simulate the dynamical thermohydraulic behavior of a single-effect absorption machine in transient conditions, being im...

  5. Identifying the optimal supply temperature in district heating networks - A modelling approach

    DEFF Research Database (Denmark)

    Mohammadi, Soma; Bojesen, Carsten

    2014-01-01

    The number of low-energy and energy renovated buildings with considerably low heating demand has been continuously increasing in recent years. Combined with utilizing low temperature sources, this development raises the necessity of introducing a new generation of District Heating [DH] Systems...... of this study is to develop a model for thermo-hydraulic calculation of low temperature DH system. The modelling is performed with emphasis on transient heat transfer in pipe networks. The pseudo-dynamic approach is adopted to model the District Heating Network [DHN] behaviour which estimates the temperature...

  6. Experimental investigations at the GENEVA passive residual heat removal test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cloppenborg, Tim; Schuster, Christoph; Hurtado, Antonio [Technische Univ. Dresden (Germany). Professur fuer Wasserstoff- und Kernenergietechnik

    2014-07-01

    Phenomena of heat transfer system at low driving forces - mainly the transition zone between single phase and two phase heat transfer - is of high interest for several technical applications. Passive safety systems of advanced nuclear reactor concepts and operation of concentrated solar power systems are only two examples. The GENEVA natural circulation test facility was established for generic investigations of thermohydraulic impact factors on natural circulation residual heat removal systems at the Professorship of Hydrogen- and Nuclear Energy Technology, TU Dresden in 2013. (orig.)

  7. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  8. Comparisons of Wilks’ and Monte Carlo Methods in Response to the 10CFR50.46(c) Proposed Rulemaking

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-10-01

    The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10CFR50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently nearly all the fuel rods in a reactor core need to be analyzed under LOCA conditions to demonstrate compliance to the safety limits. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of the limiting rods as well as to quantify the safety margins under LOCA conditions. With the new analysis method presented in this work, the limiting transient case and the limiting rods can be easily identified to quantify the safety margins in response to the proposed new rulemaking. In this work, the best-estimate plus uncertainty (BEPU) analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. Both the direct Monte Carlo method and the Wilks’ nonparametric statistical method can be used to perform uncertainty quantification. Wilks’ method has become the de-facto industry standard to perform uncertainty quantification in BEPU LOCA analyses. Despite its widespread adoption by the industry, the use of small sample sizes to infer statement of compliance to the existing 10CFR50.46 rule, has been a major cause of unrealized operational margin in today’s BEPU methods. Moreover the debate on the proper interpretation of the Wilks’ theorem in the context of safety analyses is not fully resolved yet, even more than two decades after its introduction in the frame of safety analyses in the nuclear industry. This represents both a regulatory

  9. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  10. Modeling and Simulation of Turbulent Flows through a Solar Air Heater Having Square-Sectioned Transverse Rib Roughness on the Absorber Plate

    Directory of Open Access Journals (Sweden)

    Anil Singh Yadav

    2013-01-01

    Full Text Available Solar air heater is a type of heat exchanger which transforms solar radiation into heat energy. The thermal performance of conventional solar air heater has been found to be poor because of the low convective heat transfer coefficient from the absorber plate to the air. Use of artificial roughness on a surface is an effective technique to enhance the rate of heat transfer. A CFD-based investigation of turbulent flow through a solar air heater roughened with square-sectioned transverse rib roughness has been performed. Three different values of rib-pitch (P and rib-height (e have been taken such that the relative roughness pitch (P/e=14.29 remains constant. The relative roughness height, e/D, varies from 0.021 to 0.06, and the Reynolds number, Re, varies from 3800 to 18,000. The results predicted by CFD show that the average heat transfer, average flow friction, and thermohydraulic performance parameter are strongly dependent on the relative roughness height. A maximum value of thermohydraulic performance parameter has been found to be 1.8 for the range of parameters investigated. Comparisons with previously published work have been performed and found to be in excellent agreement.

  11. Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

    Science.gov (United States)

    Koltashev, D. A.; Stakhanova, A. A.

    2017-01-01

    This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

  12. System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment

    Science.gov (United States)

    Usov, E. V.; Butov, A. A.; Dugarov, G. A.; Kudasov, I. G.; Lezhnin, S. I.; Mosunova, N. A.; Pribaturin, N. A.

    2017-07-01

    The system of equations from a two-fluid model is widely used in modeling thermohydraulic processes during accidents in nuclear reactors. The model includes conservation equations governing the balance of mass, momentum, and energy in each phase of the coolant. The features of heat and mass transfer, as well as of mechanical interaction between phases or with the channel wall, are described by a system of closing relations. Properly verified foreign and Russian codes with a comprehensive system of closing relations are available to predict processes in water coolant. As to the sodium coolant, only a few open publications on this subject are known. A complete system of closing relations used in the HYDRA-IBRAE/LM/V1 thermohydraulic code for calculation of sodium boiling in channels of power equipment is presented. The selection of these relations is corroborated on the basis of results of analysis of available publications with an account taken of the processes occurring in liquid sodium. A comparison with approaches outlined in foreign publications is presented. Particular attention has been given to the calculation of the sodium two-phase flow boiling. The flow regime map and a procedure for the calculation of interfacial friction and heat transfer in a sodium flow with account taken of high conductivity of sodium are described in sufficient detail. Correlations are presented for calculation of heat transfer for a single-phase sodium flow, sodium flow boiling, and sodium flow boiling crisis. A method is proposed for prediction of flow boiling crisis initiation.

  13. Reconstruction of the limit cycles by the delays method; Reconstruccion de ciclos limite por el metodo de los retardos

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Calleros M, G. [CFE, CNLV, Alto Lucero, Veracruz (Mexico)]. e-mail: rcd@nuclear.inin.mx

    2003-07-01

    The boiling water reactors (BWRs) are designed for usually to operate in a stable-lineal regime. In a limit cycle the behavior of the one system is no lineal-stable. In a BWR, instabilities of nuclear- thermohydraulics nature can take the reactor to a limit cycle. The limit cycles should to be avoided since the oscillations of power can cause thermal fatigue to the fuel and/or shroud. In this work the employment of the delays method is analyzed for its application in the detection of limit cycles in a nuclear power plant. The foundations of the method and it application to power signals to different operation conditions are presented. The analyzed signals are: to steady state, nuclear-thermohydraulic instability, a non linear transitory and, finally, failure of a controller plant . Among the main results it was found that the delays method can be applied to detect limit cycles in the power monitors of the BWR reactors. It was also found that the first zero of the autocorrelation function is an appropriate approach to select the delay in the detection of limit cycles, for the analyzed cases. (Author)

  14. Saphyr: a code system from reactor design to reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)

    2003-07-01

    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.

  15. Experimental study on fluid mixing in a fuel subassembly of a fast reactor. Temperature field around heated pin with cross flow

    Energy Technology Data Exchange (ETDEWEB)

    Miyakoshi, Hiroyuki; Kamide, Hideki; Tanaka, Masaaki; Yamamoto, Kazuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-03-01

    High burnup of the core is one of means to reduce the cost of a fast reactor and fuel cycle system. However, it is not enough to investigate thermohydraulics in the core, in which fuel and wrapper tube are deformed due to irradiation under high burnup condition. In this study, sodium experiment was performed to investigate fluid mixing in a wire-wrapped 37-pin subassembly model, which had local blockage and cross flow around the blockage. Such cross flow is one of elements of thermohydraulics in a deformed subassembly. The experimental results is useful to develop numerical simulation method for the deformed subassembly. Seven pins, each had different relative position to the blockage, were heated individually in the experiments. Temperature field in the subassembly was measured. Influences of the flow rate and heater power were also examined. A horizontal cross flow occurred in upstream region toward the blockage. It was observed that the temperature field was influenced by this cross flow. The measured temperature field showed that there was a bypass flow around the blockage, which flowed toward the center of subassembly. The cross flow due to the bypass flow reached the 3rd row of pins from the blockage. The swirl flow, resulted from the spacer wire, also influenced the temperature field. The obtained experimental data will be used to develop and verify a numerical simulation method for a deformed fuel subassembly. (author)

  16. Shutdown margin for high conversion BWRs operating in Th-{sup 233}U fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Shaposhnik, Y., E-mail: shaposhy@bgu.ac.il [NRCN – Nuclear Research Center Negev, POB 9001, Beer Sheva 84190 (Israel); Department of Nuclear Engineering, Ben-Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Elias, E. [Faculty of Mechanical Engineering, Technion – Israel Institute of Technology, Technion City 32000, Haifa (Israel)

    2014-09-15

    Highlights: • BWR core operating in a closed self-sustainable Th-{sup 233}U fuel cycle. • Shutdown Margin in Th-RBWR design. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal–hydraulic analysis includes MCPR observation. - Abstract: Several reactivity control system design options are explored in order to satisfy shutdown margin (SDM) requirements in a high conversion BWRs operating in Th-{sup 233}U fuel cycle (Th-RBWR). The studied core has an axially heterogeneous fuel assembly structure with a single fissile zone “sandwiched” between two fertile blanket zones. The utilization of an originally suggested RBWR Y-shape control rod in Th-RBWR is shown to be insufficient for maintaining adequate SDM to balance the high negative reactivity feedbacks, while maintaining fuel breeding potential, core power rating, and minimum Critical Power Ratio (CPR). Implementation of alternative reactivity control materials, reducing axial leakage through non-uniform enrichment distribution, use of burnable poisons, reducing number of pins as well as increasing pin diameter are also shown to be incapable of meeting the SDM requirements. Instead, an alternative assembly design, based on Rod Cluster Control Assembly with absorber rods was investigated. This design matches the reference ABWR core power and has adequate shutdown margin. The new concept was modeled as a single three-dimensional fuel assembly having reflective radial boundaries, using the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules.

  17. Procedure of calculation of the spatial distribution of temperatures and heat fluxes in the steam generator of a nuclear power installation with an RBEC fast-neutron reactor

    Science.gov (United States)

    Frolov, A. A.; Sedov, A. A.

    2016-08-01

    A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.

  18. The 400W at 1.8K Test Facility at CEA-Grenoble

    Science.gov (United States)

    Roussel, P.; Girard, A.; Jager, B.; Rousset, B.; Bonnay, P.; Millet, F.; Gully, P.

    2006-04-01

    A new test facility with a cooling capacity respectively of 400W at 1.8K or 800W at 4.5K, is now under nominal operation in SBT (Low Temperature Department) at CEA Grenoble. It has been recently used for thermohydraulic studies of two phase superfluid helium in autumn 2004. In the near future, this test bench will allow: - to test industrial components at 1.8K (magnets, cavities of accelerators) - to continue the present studies on thermohydraulics of two phase superfluid helium - to develop and simulate new cooling loops for ITER Cryogenics, and other applications such as high Reynolds number flows This new facility consists of a cold box connected to a warm compressor station (one subatmospheric oil ring pump in series with two screw compressors). The cold box, designed by AIR LIQUIDE, comprises two centrifugal cold compressors, a cold turbine, a wet piston expander, counter flow heat exchangers and two phase separators at 4.5K and 1.8K. The new facility uses a Programmable Logic Controller (PLC) connected to a bus for the measurements. The design is modular and will allow the use of saturated fluid flow (two phase flow at 1.8K or 4.5K) or single phase fluid forced flow. Experimental results and cooling capacity in different operation modes are detailed.

  19. Boiling process in oil coolers on porous elements

    Directory of Open Access Journals (Sweden)

    Genbach Alexander A.

    2016-01-01

    Full Text Available Holography and high-speed filming were used to reveal movements and deformations of the capillary and porous material, allowing to calculate thermo-hydraulic characteristics of boiling liquid in the porous structures. These porous structures work at the joint action of capillary and mass forces, which are generalised in the form of dependences used in the calculation for oil coolers in thermal power plants (TPP. Furthermore, the mechanism of the boiling process in porous structures in the field of mass forces is explained. The development process of water steam formation in the mesh porous structures working at joint action of gravitational and capillary forces is investigated. Certain regularities pertained to the internal characteristics of boiling in cells of porous structure are revealed, by means of a holographic interferometry and high-speed filming. Formulas for calculation of specific thermal streams through thermo-hydraulic characteristics of water steam formation in mesh structures are obtained, in relation to heat engineering of thermal power plants. This is the first calculation of heat flow through the thermal-hydraulic characteristics of the boiling process in a reticulated porous structure obtained by a photo film and holographic observations.

  20. 1996 outstanding facts; Faits marquants 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    This progress report of the Direction of Studies and Research (DER) of Electricite de France (EdF) reports on some outstanding studies carried out during the year 1996 and concerning: new applications of electric power (thermal comfort, heating floors, building diagnosis, energy management, customers communicating interfaces, services, air conditioning, off-peak tariffs, power demand mastery in the industry, infrared dryers for paper industry); production means (automatic systems for nuclear power plants operation, management of reactors shutdown schedules for refuelling operations, optimization of fuel loading patterns for PWRs, neutronic and thermohydraulic computer codes for steam pipes rupture accidents, thermo-hydraulic modeling of the confinement building during post-accidental situation, 3D numerical simulation of overpressures inside reactor valves and of vortex and two-phase flow inside auxiliary pipes, control of vibrating fatigue of pipe nozzles, qualification of the adjustable speed drives of the Gravelines` cooling pumps, 3D analysis of new steam turbine designs, identification of bi-metal welding surface defects, a simple method for the evaluation of in-service fatigue of components, the probabilistic dimensioning of safety coefficients, the modeling of thermo-hydro-mechanical coupling of geo-materials for radioactive wastes storage, the supply of isolated sites using renewable energies); environment protection (batteries for electric-powered vehicles, modeling of atmospheric reactive flows, chlorination of the Dampierre`s cooling circuits for pathogen amoebas elimination, in-situ treatment of PCBs isolated transformers); and development and exploitation of materials for power networks. (J.S.).

  1. Drug: D00754 [KEGG MEDICUS

    Lifescience Database Archive (English)

    Full Text Available D00754 Drug Thalidomide (JAN/USP/INN); Thalomid (TN); Thaled (TN) C13H10N2O4 258.06... Other Antitumors D00754 Thalidomide (JAN/USP/INN) Anatomical Therapeutic Chemical (ATC) classification [BR:...ther immunosuppressants L04AX02 Thalidomide D00754 Thalidomide (JAN/USP/INN) USP ...drug classification [BR:br08302] Antineoplastics Antiangiogenic Agents Thalidomide D00754 Thalidomide (JAN/U...SP/INN) Antineoplastics [BR:br08308] Biologic response modifiers Nonspecific immunomodulation Thalidomide [ATC:L04AX02] D00754 Thalid

  2. Uff… muchos tangrams para una misma aula de matemáticas

    OpenAIRE

    Martínez-Santaolalla, Manuel José; Molina, Marta; Peñas, María; María C. Cañadas; Gallardo, Sandra

    2007-01-01

    En este trabajo presentamos diversos tipos de Tangrams y algunas actividades en las que se utiliza este puzzle como recurso didáctico en el aula de matemáticas de Educación Secundaria. Se trabajan gran variedad de nociones geométricas como la semejanza de figuras, igualdad de lados, perímetro, área, simetría, Teorema de Pitágoras y de Thales, medidas aproximadas y exactas, fracciones, medidas de longitud y superficie, razón de semejanza, entre otras. Se proponen tareas de reconocimiento, cons...

  3. Advances in Navigation Sensors and Integration Technology (Les avancees en matiere de capteurs de navigation et de technologies d’integration)

    Science.gov (United States)

    2004-02-01

    also referred to as a Foucault pendulum gyroscope. Rate about the z-axis (i.e., about the vertical post) is detected by the Coriolis acceleration...paper, DGA/STTC/DTGN: Eric PLESKA MBDA F: Jacky GROSSET SAGEM SA: Jean Michel CARON THALES Avionics; Charles DUSSURGEY CEA-LETI...Gilles DELAPIERRE CEM2/Montpellier: André BOYER IEF: Alain BOSSEBOEUF LPMO: Michel de la BACHELERIE ONERA: Pierre TOUBOUL ²²²²²²²²²²²² RTO

  4. CRITICAL THINKING ABOUT TRUTH IN TEACHING (An Educational Philosophy Perspective

    Directory of Open Access Journals (Sweden)

    Eko Ariwidodo

    2008-07-01

    Full Text Available Pemikiran kritis harus disesuaikan dengan kebenaran yang sesuai dengan kapasitasnya. Boleh jadi beberapa pertimbangan tentang pemikiran dari tujuan pendidikan harus lebih dikhususkan sesuai dengan bidangnya, serta lebih berharga, yang selanjutnya disebut kritis, berhubungan dengan pertimbangan standar epistemic yang menginformasikan tentang filsafat Barat melalui Thales, lebih jelasnya sejak Socrates dan Plato. Tulisan ini akan lebih memfokuskan pada pertanyaan tentang mengajar kebenaran, pada saat pemikiran terbentuk di dalam suatu epistemologi kehidupan. Hal tersebut akan disertai pembelajaran tentang kebenaran, yang hanya terlihat sama kritisnya dengan pemikiran yang didapatkan.

  5. Un análisis del tratamiento de la semejanza en los documentos oficiales y textos escolares de matemáticas en la segunda mitad del siglo XX

    OpenAIRE

    Escudero Pérez, Isabel

    2005-01-01

    Este trabajo centra su atención en el tratamiento de la semejanza y el teorema de Thales en los documentos curriculares oficiales y en los libros de texto de matemáticas correspondientes a las edades de 11-16 años durante los últimos cincuenta años. La distinción de tres momentos en la evolución histórica de la semejanza ha permitido identifi car tres aproximaciones al concepto cuando se considera como objeto de enseñanza. Estas aproximaciones y la forma de establecer las relaciones entre sem...

  6. Part and whole in physics: An introduction

    Science.gov (United States)

    Healey, Richard; Uffink, Jos

    2013-02-01

    Natural philosophy began in ancient Ionia as thinkers such as Thales, Anaxagoras and Democritus set out to establish the composition of the world using reason and observation. Although the title Newton chose for his major work located it firmly in that tradition, the technical content of the Principia already made it largely unintelligible to the (medically trained) philosopher Locke. Propelled by the enormous growth of knowledge during and after what the historian Stephen Brush called the second scientific revolution (which he dated 1800-1950), physics and philosophy have each become increasingly professionalized and specialized, as readers of this journal are well aware.

  7. Quantum revivals of Morse oscillators and Farey-Ford geometry

    Science.gov (United States)

    Li, Alvason Zhenhua; Harter, William G.

    2015-07-01

    Analytical eigensolutions for Morse oscillators are used to investigate quantum resonance and revivals and show how Morse anharmonicity affects revival times. A minimum semi-classical Morse revival time Tmin-rev found by Heller is related to a complete quantum revival time Trev using a quantum deviation δN parameter that in turn relates Trev to the maximum quantum beat period Tmax-beat. Also, number theory of Farey and Thales-circle geometry of Ford is shown to elegantly analyze and display fractional revivals. Such quantum dynamical analysis may have applications for spectroscopy or quantum information processing and computing.

  8. A TT&C Performance Simulator for Space Exploration and Scientific Satellites - Architecture and Applications

    Science.gov (United States)

    Donà, G.; Faletra, M.

    2015-09-01

    This paper presents the TT&C performance simulator toolkit developed internally at Thales Alenia Space Italia (TAS-I) to support the design of TT&C subsystems for space exploration and scientific satellites. The simulator has a modular architecture and has been designed using a model-based approach using standard engineering tools such as MATLAB/SIMULINK and mission analysis tools (e.g. STK). The simulator is easily reconfigurable to fit different types of satellites, different mission requirements and different scenarios parameters. This paper provides a brief description of the simulator architecture together with two examples of applications used to demonstrate some of the simulator’s capabilities.

  9. XCAN — A coherent amplification network of femtosecond fiber chirped-pulse amplifiers

    Science.gov (United States)

    Daniault, L.; Bellanger, S.; Le Dortz, J.; Bourderionnet, J.; Lallier, É.; Larat, C.; Antier-Murgey, M.; Chanteloup, J.-C.; Brignon, A.; Simon-Boisson, C.; Mourou, G.

    2015-10-01

    The XCAN collaboration program between the Ecole Polytechnique and Thales aims at developing a laser system based on the coherent combination of several tens of laser beams produced through a network of amplifying optical fibers [1]. As a first step this project aspires to demonstrate the scalability of a combining architecture in the femtosecond regime providing high peak power with high repetition rate and high efficiency. The initial system will include 61 individual phased beams aimed to provide 10 mJ, 350 fs pulses at 50 kHz.

  10. A robust mission concept for a low-cost Ceres Plume Sample Return

    Science.gov (United States)

    Poncy, J.; Fontdecaba, J.; Couzin, P.

    2014-04-01

    The recent discovery of ejecta from dwarf planet Ceres by scientists [1] using ESA's Herschel telescope provides for a golden opportunity for a low cost sample return mission for very high value science return. NASA's mission Dawn will arrive at Ceres in 2015 and pave the way for future missions to Ceres. Thales Alenia Space presents here an original short-duration low-cost mission concept that provides for two low altitude fly-by's of Ceres and returns samples from the plumes to the Earth. Mission parameters are discussed and preliminary assessed in view of maximizing mission success.

  11. International Project Management Committee: Overview and Activities

    Science.gov (United States)

    Hoffman, Edward

    2010-01-01

    This slide presentation discusses the purpose and composition of the International Project Management Committee (IMPC). The IMPC was established by members of 15 space agencies, companies and professional organizations. The goal of the committee is to establish a means to share experiences and best practices with space project/program management practitioners at the global level. The space agencies that are involved are: AEB, DLR, ESA, ISRO, JAXA, KARI, and NASA. The industrial and professional organizational members are Comau, COSPAR, PMI, and Thales Alenia Space.

  12. The nature of water: Greek thought from Homer to Acusilaos.

    Science.gov (United States)

    De Santo, Rosa Maria; Bisaccia, Carmela; Cirillo, Massimo; Pollastro, Rosa Maria; Raiola, Ilaria; De Santo, Luca Salvatore

    2009-01-01

    Greek philosophy finds its roots in the myth of Homer's and Hesiod's poems and especially in Orphism which introduced the concept of a soul separated from the body with an independent principle, psiche (soul), to be rewarded or punished after death. Orphism was an important step in Greek culture. It introduced the divine into man, the soul which does not die with the body and reincarnates. From Orphism started the need of rituals capable of separating the spirit from the body. From Homer to Acusilaos, water was a very important element which connected humans and gods, long before Thales of Miletus defined it the arche.

  13. Towards telecommunication payloads with photonic technologies

    Science.gov (United States)

    Vono, S.; Di Paolo, G.; Piccinni, M.; Pisano, A.; Sotom, M.; Aveline, M.; Ginestet, P.

    2017-11-01

    In the last decade, Thales Alenia Space has put a lot of its research effort on Photonic Technologies for Space Application with the aim to offer the market satellite telecommunication systems better performance and lower costs. This research effort has been concentrated on several activities, some of them sponsored by ESA. Most promising applications refer to Payload Systems. In particular, photonic payload applications have been investigated through the following two ESA studies: Artes-1 "Next Generation Telecommunication Payloads based on Photonic Technologies" and Artes-5 "OWR - Optical Wideband Receiver" activities.

  14. Review of the Safety Concern Related to CANDU Moderator Temperature Distribution and Status of KAERI Moderator Circulation Test (MCT) Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Bo W.; Kim, Hyoung T. [Severe Accident and PHWR Safety Research Division, Daejeon (Korea, Republic of); Kim, Tongbeum [University of the Witwatersrand, Johannesburg (South Africa); Im, Sunghyuk [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep threshold temperature and no further deformation is expected. Consequently, a sufficient condition to ensure fuel channel integrity following a large LOCA, is the avoidance of sustained calandria tubes dryout. If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The subcooling requirements are determined from a set of experiments known as fuel channel contact experiments. The difference between available subcooling and required subcooling is called subcooling margins. The moderator flow circulation patterns are complicated slow flows that significantly vary from buoyancy dominated to inertia dominated patterns. Accurate predictions of flow patterns are essential for accurate calculation of moderator temperature distributions and the related moderator subcooling. Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep

  15. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  16. The CSNI/PWG-1 international task group on ECCS reliability

    Energy Technology Data Exchange (ETDEWEB)

    Sandervag, O.; Riekert, T.; Serkiz, A.; Hyvarinen, J.

    1996-03-01

    A steam line loss-of-coolant accident (LOCA) occurred when a safety relief valve inadvertently opened in the Barseback-2 nuclear power plant. The steam jet stripped fibrous insulation from adjacent pipework. Part of that insulation debris was transported to the wetwell pool and clogged the intake strainers for the drywell spray system after about one hour. Although the incident in itself was not very serious, it revealed a weakness in the defense-in-depth concept which under other circumstances could have led to failure of the emergency core cooling system (ECCS) to provide water to the core. Before the Barseback-2 LOCA, international regulators of nuclear power plants and the nuclear power plant industry had considered safety questions related to strainer clogging as resolved. Many European countries had followed the guidance for strainers in pressurized water reactors (PWRs) contained in United States Nuclear Regulatory Commission`s (USNRC) Regulatory Guide 1.82, Water Sources for Long Term Recirculation Cooling Following a Loss-of-Coolant Accident, 1974. However, data obtained from European experimental programs carried out in the late seventies to determine the performance of strainers indicated that this guide was not adequate. In addition, Swedish plant owners had used this guidance to judge performance of emergency core cooling systems (ECCS) in their plants. Analyses at that time had indicated that strainer clogging, if occurring at all, would at least not occur during the first ten hours after a LOCA. Since operation of the ECCS would be needed for a long time, backflushing capabilities and monitors of pressure drop across the strainers were installed in older Swedish BWR plants with small strainer areas. These actions were judged to be adequate compliance with the revised USNRC Regulatory Guide 1.82, Rev. 1, issued in 1985. Safety questions related to strainer clogging were considered to have been resolved until the incident happened in Barseback-2.

  17. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  18. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  19. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station; Analisis de eventos internos para la Unidad 1 de la Central Nucleolelectrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1993-07-01

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  20. Feasibility Study on Thimble Plug Removal for Westinghouse Type PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Sup; Lee, Jae Yong; Yoon, Duk Joo; Jun, Hwang Yong; Kim, Yoon Ho [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    1. Abstract of Thimble Plug Removal- Thimble Plug Removal from the core increase core bypass flow few percent and may reduce DNBR Margin 2{approx}3%. In this feasibility study, the following analyses were performed in terms of the best estimate flow, bypass flow, DNBR margin etc. 2. Area of analysis and evaluations (a). Thermal Hydraulic (b). PCWG (c). Nuclear Design (d). Rod Performance (e). mechanical Design (f). Transient Analysis (g). LOCA Analysis. 3. Evaluation of Economic and Licensing 4. Detail analysis and design were performed for Youngkwang unit 1 as a sample plant. (author). 68 refs., figs.