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Sample records for core configuration design

  1. Design configuration of GCFR core assemblies

    International Nuclear Information System (INIS)

    LaBar, M.P.; Lee, G.E.; Meyer, R.J.

    1980-05-01

    The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the same assemblies would be used in a commercial plant

  2. Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors

    International Nuclear Information System (INIS)

    Hedayat, Afshin; Davilu, Hadi; Barfrosh, Ahmad Abdollahzadeh; Sepanloo, Kamran

    2009-01-01

    To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core configuration design. Suggested parameters and conditions are considered as two separate fitness functions composed of two main objectives and two penalty functions. This is a constrained and combinatorial type of a multi-objective optimization problem. In this paper, a fast and effective hybrid artificial intelligence algorithm is introduced and developed to reach a Pareto optimal set. The hybrid algorithm is composed of a fast and elitist multi-objective genetic algorithm (GA) and a fast fitness function evaluating system based on the cascade feed forward artificial neural networks (ANNs). A specific GA representation of core configuration and also special GA operators are introduced and used to overcome the combinatorial constraints of this optimization problem. A software package (Core Pattern Calculator 1) is developed to prepare and reform required data for ANNs training and also to revise the optimization results. Some practical test parameters and conditions are suggested to adjust main parameters of the hybrid algorithm. Results show that introduced ANNs can be trained and estimate selected core parameters of a research reactor very quickly. It improves effectively optimization process. Final optimization results show that a uniform and dense diversity of Pareto fronts are gained over a wide range of fitness function values. To take a more careful selection of Pareto optimal solutions, a revision system is introduced and used. The revision of gained Pareto optimal set is performed by using developed software package. Also some secondary operational

  3. Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aut.ac.i [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of); Davilu, Hadi [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Barfrosh, Ahmad Abdollahzadeh [Department of Computer Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Sepanloo, Kamran [Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of)

    2009-12-15

    To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core configuration design. Suggested parameters and conditions are considered as two separate fitness functions composed of two main objectives and two penalty functions. This is a constrained and combinatorial type of a multi-objective optimization problem. In this paper, a fast and effective hybrid artificial intelligence algorithm is introduced and developed to reach a Pareto optimal set. The hybrid algorithm is composed of a fast and elitist multi-objective genetic algorithm (GA) and a fast fitness function evaluating system based on the cascade feed forward artificial neural networks (ANNs). A specific GA representation of core configuration and also special GA operators are introduced and used to overcome the combinatorial constraints of this optimization problem. A software package (Core Pattern Calculator 1) is developed to prepare and reform required data for ANNs training and also to revise the optimization results. Some practical test parameters and conditions are suggested to adjust main parameters of the hybrid algorithm. Results show that introduced ANNs can be trained and estimate selected core parameters of a research reactor very quickly. It improves effectively optimization process. Final optimization results show that a uniform and dense diversity of Pareto fronts are gained over a wide range of fitness function values. To take a more careful selection of Pareto optimal solutions, a revision system is introduced and used. The revision of gained Pareto optimal set is performed by using developed software package. Also some secondary operational

  4. Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

    International Nuclear Information System (INIS)

    Fox, J.N.; Wei, B.C.

    1978-02-01

    Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

  5. Viscous Design of TCA Configuration

    Science.gov (United States)

    Krist, Steven E.; Bauer, Steven X. S.; Campbell, Richard L.

    1999-01-01

    The goal in this effort is to redesign the baseline TCA configuration for improved performance at both supersonic and transonic cruise. Viscous analyses are conducted with OVERFLOW, a Navier-Stokes code for overset grids, using PEGSUS to compute the interpolations between overset grids. Viscous designs are conducted with OVERDISC, a script which couples OVERFLOW with the Constrained Direct Iterative Surface Curvature (CDISC) inverse design method. The successful execution of any computational fluid dynamics (CFD) based aerodynamic design method for complex configurations requires an efficient method for regenerating the computational grids to account for modifications to the configuration shape. The first section of this presentation deals with the automated regridding procedure used to generate overset grids for the fuselage/wing/diverter/nacelle configurations analysed in this effort. The second section outlines the procedures utilized to conduct OVERDISC inverse designs. The third section briefly covers the work conducted by Dick Campbell, in which a dual-point design at Mach 2.4 and 0.9 was attempted using OVERDISC; the initial configuration from which this design effort was started is an early version of the optimized shape for the TCA configuration developed by the Boeing Commercial Airplane Group (BCAG), which eventually evolved into the NCV design. The final section presents results from application of the Natural Flow Wing design philosophy to the TCA configuration.

  6. Minimars configuration design

    International Nuclear Information System (INIS)

    Lousteau, D.C.

    1986-01-01

    Viewgraph comprise the report. Aspects of design which are discussed are: end cell coil arrangement, neutral beam and rf injection, load definition on end cell coils, and intracoil structure for all coils

  7. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  8. Turbofan Engine Core Compartment Vent Aerodynamic Configuration Development Methodology

    Science.gov (United States)

    Hebert, Leonard J.

    2006-01-01

    This paper presents an overview of the design methodology used in the development of the aerodynamic configuration of the nacelle core compartment vent for a typical Boeing commercial airplane together with design challenges for future design efforts. Core compartment vents exhaust engine subsystem flows from the space contained between the engine case and the nacelle of an airplane propulsion system. These subsystem flows typically consist of precooler, oil cooler, turbine case cooling, compartment cooling and nacelle leakage air. The design of core compartment vents is challenging due to stringent design requirements, mass flow sensitivity of the system to small changes in vent exit pressure ratio, and the need to maximize overall exhaust system performance at cruise conditions.

  9. Analysis of RA-8 critical facility core in some configurations

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  10. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  11. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  12. Design and safety aspect of lead and lead-bismuth cooled long-life small safe fast reactors for various core configurations

    International Nuclear Information System (INIS)

    Zaki, S.; Sekimoto, Hiroshi

    1995-01-01

    Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ∼2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25-40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient. (author)

  13. LMFBR core design analysis

    International Nuclear Information System (INIS)

    Cho, M.; Yang, J.C.; Yoh, K.C.; Suk, S.D.; Soh, D.S.; Kim, Y.M.

    1980-01-01

    The design parameters of a commercial-scale fast breeder reactor which is currently under construction by regeneration of these data is preliminary analyzed. The analysis of nuclear and thermal characteristics as well as safety features of this reactor is emphasized. And the evaluation of the initial core mentioned in the system description is carried out in the areas of its kinetics and control system, and, at the same time, the flow distribution of sodium and temperature distribution of the initial FBR core system are calculated. (KAERI INIS Section)

  14. Statistical core design

    International Nuclear Information System (INIS)

    Oelkers, E.; Heller, A.S.; Farnsworth, D.A.; Kearfott, K.J.

    1978-01-01

    The report describes the statistical analysis of DNBR thermal-hydraulic margin of a 3800 MWt, 205-FA core under design overpower conditions. The analysis used LYNX-generated data at predetermined values of the input variables whose uncertainties were to be statistically combined. LYNX data were used to construct an efficient response surface model in the region of interest; the statistical analysis was accomplished through the evaluation of core reliability; utilizing propagation of the uncertainty distributions of the inputs. The response surface model was implemented in both the analytical error propagation and Monte Carlo Techniques. The basic structural units relating to the acceptance criteria are fuel pins. Therefore, the statistical population of pins with minimum DNBR values smaller than specified values is determined. The specified values are designated relative to the most probable and maximum design DNBR values on the power limiting pin used in present design analysis, so that gains over the present design criteria could be assessed for specified probabilistic acceptance criteria. The results are equivalent to gains ranging from 1.2 to 4.8 percent of rated power dependent on the acceptance criterion. The corresponding acceptance criteria range from 95 percent confidence that no pin will be in DNB to 99.9 percent of the pins, which are expected to avoid DNB

  15. Automated Core Design

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2005-01-01

    Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process

  16. Design review of the INTOR mechanical configuration

    International Nuclear Information System (INIS)

    Brown, T.G.

    1981-01-01

    The INTOR conceptual design has been carried out by design teams working in the home countries with periodic workshop sessions in Vienna to review the ongoing work and to make decisions on the evolving design. The decisions taken at each workshop session were then incorporated into each national design activity, so that the four national design contributions would progressively converge toward a single design with increasingly greater detail. This paper defines the final INTOR configuration that has evolved during the conceptual design phase, defining the major system design alternatives that were considered and the rationale for selecting the final system configuration

  17. Design configuration for Cernavoda - 2

    International Nuclear Information System (INIS)

    Lindsay, R.I.; Keil, H.; Hapwood, J.M.; Hum, J.

    1998-01-01

    The Cernavoda - 2 NPP project is based on a repeat of the recently completed plant, Cernavoda - 1 NPP which has been operating well, and the other CANDU 6 units in Canada, Argentina and Korea, which have had an excellent operating record. The reference plant design for Cernavoda - 2 is Cernavoda - 1 with appropriate design enhancements incorporating lessons learned from CANDU 6 and other nuclear operations, and including appropriate design enhancements from the most recent CANDU 6 projects at Wolsung in Korea, and Qinshan in China, while recognizing the significant degree of project completion. The results of this combination of proven design with systematic design feedback, will be a unit which combines reliability supported by the many years of successful operation of CANDU 6 units, together with a significant number of design enhancements. (authors)

  18. Evolution of the Configuration Database Design

    International Nuclear Information System (INIS)

    Salnikov, A.

    2006-01-01

    The BABAR experiment at SLAC successfully collects physics data since 1999. One of the major parts of its on-line system is the configuration database which provides other parts of the system with the configuration data necessary for data taking. Originally the configuration database was implemented in the Objectivity/DB ODBMS. Recently BABAR performed a successful migration of its event store from Objectivity/DB to ROOT and this prompted a complete phase-out of the Objectivity/DB in all other BABAR databases. It required the complete redesign of the configuration database to hide any implementation details and to support multiple storage technologies. In this paper we describe the process of the migration of the configuration database, its new design, implementation strategy and details

  19. Core design methods for advanced LMFBRs

    International Nuclear Information System (INIS)

    Chandler, J.C.; Marr, D.R.; McCurry, D.C.; Cantley, D.A.

    1977-05-01

    The multidiscipline approach to advanced LMFBR core design requires an iterative design procedure to obtain a closely-coupled design. HEDL's philosophy requires that the designs should be coupled to the extent that the design limiting fuel pin, the design limiting duct and the core reactivity lifetime should all be equal and should equal the fuel residence time. The design procedure consists of an iterative loop involving three stages of the design sequence. Stage 1 consists of general mechanical design and reactor physics scoping calculations to arrive at an initial core layout. Stage 2 consists of detailed reactor physics calculations for the core configuration arrived at in Stage 1. Based upon the detailed reactor physics results, a decision is made either to alter the design (Stage 1) or go to Stage 3. Stage 3 consists of core orificing and detailed component mechanical design calculations. At this point, an assessment is made regarding design adequacy. If the design is inadequate the entire procedure is repeated until the design is acceptable

  20. Space shuttle configuration accounting functional design specification

    Science.gov (United States)

    1974-01-01

    An analysis is presented of the requirements for an on-line automated system which must be capable of tracking the status of requirements and engineering changes and of providing accurate and timely records. The functional design specification provides the definition, description, and character length of the required data elements and the interrelationship of data elements to adequately track, display, and report the status of active configuration changes. As changes to the space shuttle program levels II and III configuration are proposed, evaluated, and dispositioned, it is the function of the configuration management office to maintain records regarding changes to the baseline and to track and report the status of those changes. The configuration accounting system will consist of a combination of computers, computer terminals, software, and procedures, all of which are designed to store, retrieve, display, and process information required to track proposed and proved engineering changes to maintain baseline documentation of the space shuttle program levels II and III.

  1. Sustainable supply chain design: a configurational approach.

    Science.gov (United States)

    Masoumik, S Maryam; Abdul-Rashid, Salwa Hanim; Olugu, Ezutah Udoncy; Raja Ghazilla, Raja Ariffin

    2014-01-01

    Designing the right supply chain that meets the requirements of sustainable development is a significant challenge. Although there are a considerable number of studies on issues relating to sustainable supply chain design (SSCD) in terms of designing the practices, processes, and structures, they have rarely demonstrated how these components can be aligned to form an effective sustainable supply chain (SSC). Considering this gap in the literature, this study adopts the configurational approach to develop a conceptual framework that could configure the components of a SSC. In this respect, a process-oriented approach is utilized to classify and harmonize the design components. A natural-resource-based view (NRBV) is adopted to determine the central theme to align the design components around. The proposed framework presents three types of SSC, namely, efficient SSC, innovative SSC, and reputed SSC. The study culminates with recommendations concerning the direction for future research.

  2. Sustainable Supply Chain Design: A Configurational Approach

    Science.gov (United States)

    Masoumik, S. Maryam; Raja Ghazilla, Raja Ariffin

    2014-01-01

    Designing the right supply chain that meets the requirements of sustainable development is a significant challenge. Although there are a considerable number of studies on issues relating to sustainable supply chain design (SSCD) in terms of designing the practices, processes, and structures, they have rarely demonstrated how these components can be aligned to form an effective sustainable supply chain (SSC). Considering this gap in the literature, this study adopts the configurational approach to develop a conceptual framework that could configure the components of a SSC. In this respect, a process-oriented approach is utilized to classify and harmonize the design components. A natural-resource-based view (NRBV) is adopted to determine the central theme to align the design components around. The proposed framework presents three types of SSC, namely, efficient SSC, innovative SSC, and reputed SSC. The study culminates with recommendations concerning the direction for future research. PMID:24523652

  3. Feasibility neutronic conceptual design for the core configuration of a 75 kWth Aqueous Homogeneous Reactor for 99Mo production

    International Nuclear Information System (INIS)

    Milian, D.; Milian, D. E.; Rodriguez, L. P.; Salomon, J.; Cadavid, N.

    2015-01-01

    99m Tc is a very useful radioisotope, which is used in nearly 80% of all nuclear medicine procedures. 99m Tc is produced from 99 Mo decay. Since 2007 the medical community has been plagued by 99 Mo shortages due to aging reactors, such as the National Research Universal reactor in Canada and the High Flux Reactor in Petten, The Netherlands. At present, most of the world's supply of 99 Mo for medical isotope production involves the neutron fission of 235 U in multipurpose research reactors. 99 Mo mostly results from the fission reaction of 235 U targets with a fission yield about 6.1%. After irradiation in the reactor, the target is digested in acid or alkaline solutions and 99 Mo is recovered through a series of extraction (separation) and purification steps. 99 Mo production system in an Aqueous Homogeneous Reactor (AHR) offers a better method, because all of the 99 Mo can be extracted from the fuel solution. Over 30 AHRs has been built and operated around the world with 149 years of combined experience. In this paper, an AHR conceptual design using LEU (Low Enriched Uranium) is optimized to meet the South American demand for 99 Mo for the coming years. Aspect related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotope production and others are evaluated. The neutronic calculations have been performed with the well-known MCNPX computational code. A benchmarking experiments performed at the Russian Research Center 'Kurchatov Institute' in order to validate that the developed models of AHRs with MCNPX code and the available library in XSDIR, ENDF/B VI.2, are adequate for studies of aqueous fuel solutions. (Author)

  4. Overview of homogeneous versus heterogeneous core configuration trade-off studies

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1982-01-01

    The most significant development in core design trend in the U.S. LMFBR program has been the increased attention given to the heterogeneous core design concept. In recent years, numerous core configuration trade-off studies have been carried out to quantify advantages and disadvantages of the heterogeneous concept relative to the homogeneous concept, and a consensus is emerging among the U.S. core designers. It appears that the technical and economic performance differences between homogeneous and heterogeneous core designs are very small; however, the heterogeneous concept provides a definite safety/licensing advantage. he technical and economic performance comparison between homogeneous and heterogeneous core configurations is difficult to quantify. In fact, in most cases, the perceived advantages and/or disadvantages are dictated by the consistency in the comparison (optimized for one concept versus non-optimized for the other, etc.) rather than by any inherent differences. Some of the technical and economic issues relevant to the homogeneous versus heterogeneous comparison are summarized

  5. Overview of homogeneous versus heterogeneous core configuration trade-off studies

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y I [Applied Physics Division, Argonne National Laboratory, Argonne, IL (United States)

    1982-01-01

    The most significant development in core design trend in the U.S. LMFBR program has been the increased attention given to the heterogeneous core design concept. In recent years, numerous core configuration trade-off studies have been carried out to quantify advantages and disadvantages of the heterogeneous concept relative to the homogeneous concept, and a consensus is emerging among the U.S. core designers. It appears that the technical and economic performance differences between homogeneous and heterogeneous core designs are very small; however, the heterogeneous concept provides a definite safety/licensing advantage. he technical and economic performance comparison between homogeneous and heterogeneous core configurations is difficult to quantify. In fact, in most cases, the perceived advantages and/or disadvantages are dictated by the consistency in the comparison (optimized for one concept versus non-optimized for the other, etc.) rather than by any inherent differences. Some of the technical and economic issues relevant to the homogeneous versus heterogeneous comparison are summarized.

  6. Automated firewall analytics design, configuration and optimization

    CERN Document Server

    Al-Shaer, Ehab

    2014-01-01

    This book provides a comprehensive and in-depth study of automated firewall policy analysis for designing, configuring and managing distributed firewalls in large-scale enterpriser networks. It presents methodologies, techniques and tools for researchers as well as professionals to understand the challenges and improve the state-of-the-art of managing firewalls systematically in both research and application domains. Chapters explore set-theory, managing firewall configuration globally and consistently, access control list with encryption, and authentication such as IPSec policies. The author

  7. Toward full MOX core design

    International Nuclear Information System (INIS)

    Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

    1999-01-01

    This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

  8. The effect of core configuration on temperature coefficient of reactivity in IRR-1

    Energy Technology Data Exchange (ETDEWEB)

    Bettan, M.; Silverman, I.; Shapira, M.; Nagler, A. [Soreq Nuclear Research Center, Yavne (Israel)

    1997-08-01

    Experiments designed to measure the effect of coolant moderator temperature on core reactivity in an HEU swimming pool type reactor were performed. The moderator temperature coefficient of reactivity ({alpha}{sub {omega}}) was obtained and found to be different in two core loadings. The measured {alpha}{sub {omega}} of one core loading was {minus}13 pcm/{degrees}C at the temperature range of 23-30{degrees}C. This value of {alpha}{sub {omega}} is comparable to the data published by the IAEA. The {alpha}{sub {omega}} measured in the second core loading was found to be {minus}8 pcm/{degrees}C at the same temperature range. Another phenomenon considered in this study is core behavior during reactivity insertion transient. The results were compared to a core simulation using the Dynamic Simulator for Nuclear Power Plants. It was found that in the second core loading factors other than the moderator temperature influence the core reactivity more than expected. These effects proved to be extremely dependent on core configuration and may in certain core loadings render the reactor`s reactivity coefficient undesirable.

  9. Reliable core thermal design

    International Nuclear Information System (INIS)

    Amendola, A.

    1974-01-01

    The hot spot analysis is no longer limited to the calculation of a simple safety factor against overtemperature, but is now integrated in the overall design philosophy. This paper describes the development of a probabilistic method of analysis and compares it with other advanced calculation methods. Feedbacks from the analysis act: - on the nominal temperature distribution in order to satisfy the maximum temperature limit and in the same time to optimize the coolant temperature for maximum plant efficiency, and - on the specifications of manufacturing tolerances and experimental investigations in order to identify and to reduce the most important design uncertainties. Moreover the computer codes SHOSPA and THEDRA are briefly discussed. Both codes deliver the zero hot spot probability as a function of the geometrical size assumed for a ''spot''. THEDRA delivers also the expected hot spot distribution. By means of THEDRA it is possible to evaluate the pins failure expectation if the distribution of pin failures versus operating temperature is known. (author)

  10. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

    2015-10-15

    Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

  11. Monte Carlo simulation of a TRIGA source driven core configuration: Preliminary results

    International Nuclear Information System (INIS)

    Burgio, N.; Ciavola, C.; Santagata, A.

    2002-01-01

    The different core configurations with a k eff ranging from 0.93 to 0.98, and their response when driven by a pulsed neutron source were simulated with MCNP4C3 (Los Alamos - Monte Carlo N Particles). Simulation results could be considered both as preliminary check for nuclear data and a conceptual design for 'source jerk' experiments on the frame of TRIGA Accelerator Driven Experiment (TRADE) on the reactor facility of Casaccia research center. (author)

  12. Thawing of lithium in the SP-100 reactor core configuration

    International Nuclear Information System (INIS)

    Magee, P.M.; Malovrh, J.W.; REineking, W.H.

    1986-01-01

    The General Electric SP-100 Liquid Metal Reactor is designed to be launched with the lithium coolant in the reactor and primary loops frozen. Initial startup of the system in space, after a satisfactory orbit is achieved, will be accomplished by slowly increasing the power in the reactor core and using the heat generated to melt the lithium, first in the reactor, and then progressively down the primary loops. This technique significantly facilitates ground handling, reduces vibrational loads during vehicle launch and minimized the shuttle bay heat load. The challenge is to thaw the coolant and startup the system within an acceptable time without structural damage. The test results clearly demonstrate that thawing of the lithium in the SP-100 reactor core can be done rapidly without structural damage and, thus, support the selected concept of SP-100 launch with frozen lithium and thaw/startup in space

  13. Configuration management after design basis reconstitution

    International Nuclear Information System (INIS)

    Purcell, J.J.; Livingston, B.R.

    1991-01-01

    Over the last few years, Fort Calhoun station (FCS) has implemented a number of programs to enhance plant operability and readiness. The design basis document (DBD) reconstitution project was the cornerstone of this effort. Vendor manual upgrade, operating procedures upgrade, plant equipment data-base verification, equipment labeling, and warehousing improvements were also implemented as part of this improvement program. With the completion of these programs, plant documentation was current to the baselines established by each program, and a configuration management program (CMP) was established to maintain this level of accuracy throughout the remaining life of FCS. Change control throughout the organization has been reviewed and upgraded to ensure that all changes are evaluated for impact to the design bases

  14. SMART core protection system design

    International Nuclear Information System (INIS)

    Lee, J. K.; Park, H. Y.; Koo, I. S.; Park, H. S.; Kim, J. S.; Son, C. H.

    2003-01-01

    SMART COre Protection System(SCOPS) is designed with real-tims Digital Signal Processor(DSP) board and Network Interface Card(NIC) board. SCOPS has a Control Rod POSition (CRPOS) software module while Core Protection Calculator System(CPCS) consists of Core Protection Calculators(CPCs) and Control Element Assembly(CEA) Calculators(CEACs) in the commercial nuclear plant. It's not necessary to have a independent cabinets for SCOPS because SCOPS is physically very small. Then SCOPS is designed to share the cabinets with Plant Protection System(PPS) of SMART. Therefor it's very easy to maintain the system because CRPOS module is used instead of the computer with operating system

  15. MarFS-Requirements-Design-Configuration-Admin

    Energy Technology Data Exchange (ETDEWEB)

    Kettering, Brett Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Grider, Gary Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-07-08

    This document will be organized into sections that are defined by the requirements for a file system that presents a near-POSIX (Portable Operating System Interface) interface to the user, but whose data is stored in whatever form is most efficient for the type of data being stored. After defining the requirement the design for meeting the requirement will be explained. Finally there will be sections on configuring and administering this file system. More and more, data dominates the computing world. There is a “sea” of data out there in many different formats that needs to be managed and used. “Mar” means “sea” in Spanish. Thus, this product is dubbed MarFS, a file system for a sea of data.

  16. Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Stauff, N.E.; Klim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States); Fiorina, C. [Politecnico di Milano, Milan (Italy); Franceschini, F. [Westinghouse Electric Company LLC., Cranberry Township, Pennsylvania (United States)

    2013-07-01

    A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic

  17. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  18. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  19. Progress of full MOX core design in ABWR

    International Nuclear Information System (INIS)

    Izutsu, S.; Sasagawa, M.; Aoyama, M.; Maruyama, H.; Suzuki, T.

    2000-01-01

    Full MOX ABWR core design has been made, based on the MOX design concept of 8x8 bundle configuration with a large central water rod, 40 GWd/t maximum bundle exposure, and the compatibility with 9x9 high-burnup UO 2 bundles. Core performance on shutdown margin and thermal margin of the MOX-loaded core is similar to that of UO 2 cores for the range from full UO 2 core to full MOX core. Safety analyses based on its safety parameters and MOX property have shown its conformity to the design criteria in Japan. In order to confirm the applicability of the nuclear design method to full MOX cores, Tank-type Critical Assembly (TCA) experiment data have been analyzed on criticality, power distribution and β eff /l measurements. (author)

  20. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    Marth, W.; Schroeder, R.

    1983-07-01

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

  1. Dynamic studies of multiple configurations of CERN's Antiproton Decelerator Target core under proton beam impact

    CERN Document Server

    AUTHOR|(CDS)2248381

    Antiprotons, like many other exotic particles, are produced by impacting high energy proton beams onto fixed targets. At the European Organization for Nuclear Research (CERN), this is done in the Antiproton Decelerator (AD) Facility. The engineering challenges related to the design of an optimal configuration of the AD-Target system derive from the extremely high energy depositions reached in the very thin target core as a consequence of each proton beam impact. A new target design is foreseen for operation after 2021, triggering multiple R&D activities since 2013 for this purpose. The goal of the present Master Thesis is to complement these activities with analytical and numerical calculations, delving into the phenomena associated to the dynamic response of the target core. In this context, two main studies have been carried out. First, the experimental data observed in targets subjected to low intensity proton pulses was cross-checked with analytical and computational methods for modal analysis, applie...

  2. Configuration of Web services as parametric design

    NARCIS (Netherlands)

    Ten Teije, Annette; Van Harmelen, Frank; Wielinga, Bob

    2004-01-01

    The configuration of Web services is particularly hard given the heterogeneous, unreliable and open nature of the Web. Furthermore, such composite Web services are likely to be complex services, that will require adaptation for each specific use. Current approaches to Web service configuration are

  3. Configuration and structural design of Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Brown, T.G.

    1985-01-01

    Viewgraphs are presented on the configuration and structural design of the Compact Ignition Tokamak, originally presented to the US/Japan Workshop on Next Step Machine Design. Items discussed in this presentation include: PPPL 0424 ref design; MIT LITE ref design; IGNITOR 1.01 M ref design; and IGNITOR 1.08 M press configuration

  4. Conceptual design of PFBR core

    International Nuclear Information System (INIS)

    Lee, S.M.; Govindarajan, S.; Indira, R.; John, T.M.; Mohanakrishnan, P.; Shankar Singh, R.; Bhoje, S.B.

    1996-01-01

    The design options selected for the core of the 500 MWe Prototype Fast Breeder Reactor are presented. PFBR has a conventional mixed oxide fuel core of homogeneous type with two enrichment zones for power flattening and with radial and axial blankets to make the reactor self-sustaining in fissile material. Pin diameter has been selected for minimization of fissile inventory. Considerations for the choice of number of pins per subassembly, integrated versus separate axial blankets, and other pin and subassembly parameters are discussed. As the core size is moderate, no special schemes for reducing the maximum positive sodium voiding coefficient is envisages. Two independent, diverse fast acting shutdown systems working in fail-safe mode are selected. The number of absorber rods has been minimized by choosing a layout for maximum antishadow effect. Nine control and safety rods are distributed in two rods for power flattening by differential insertion. Three Diverse Safety Rods, are also provided which are normally fully withdrawn. The optimization of layout of radial and axial shielding and adequacy of flux at detector location are also discussed. (author). 2 figs

  5. Design issues for optimum solar cell configuration

    Science.gov (United States)

    Kumar, Atul; Thakur, Ajay D.

    2018-05-01

    A computer based simulation of solar cell structure is performed to study the optimization of pn junction configuration for photovoltaic action. The fundamental aspects of photovoltaic action viz, absorption, separation collection, and their dependence on material properties and deatails of device structures is discussed. Using SCAPS 1D we have simulated the ideal pn junction and shown the effect of band offset and carrier densities on solar cell performance. The optimum configuration can be achieved by optimizing transport of carriers in pn junction under effect of field dependent recombination (tunneling) and density dependent recombination (SRH, Auger) mechanisms.

  6. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  7. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  8. Design knowledge management using configuration manager

    International Nuclear Information System (INIS)

    Kang, Mu Jin; Jung, Seung Hwan

    2000-01-01

    It is known that about 15 to 40 percent of total design time is spent on retrieving information such as standard parts, handbook data, engineering equations, previous designs. This paper describes a knowledge management system for machine tool design. Product structuring, change management, and complex design knowledge management are possible through the developed system. The system can speed up the design process by making necessary data instantly available as it is needed and keeping of all the relevant design information and knowledge including individual decisions, design intentions, documents, and drawings

  9. Design Principles for Synthesizable Processor Cores

    DEFF Research Database (Denmark)

    Schleuniger, Pascal; McKee, Sally A.; Karlsson, Sven

    2012-01-01

    As FPGAs get more competitive, synthesizable processor cores become an attractive choice for embedded computing. Currently popular commercial processor cores do not fully exploit current FPGA architectures. In this paper, we propose general design principles to increase instruction throughput...

  10. Collaborative Systems Driven Aircraft Configuration Design Optimization

    OpenAIRE

    Shiva Prakasha, Prajwal; Ciampa, Pier Davide; Nagel, Björn

    2016-01-01

    A Collaborative, Inside-Out Aircraft Design approach is presented in this paper. An approach using physics based analysis to evaluate the correlations between the airframe design, as well as sub-systems integration from the early design process, and to exploit the synergies within a simultaneous optimization process. Further, the disciplinary analysis modules involved in the optimization task are located in different organization. Hence, the Airframe and Subsystem design tools are integrated ...

  11. Secure Protocol and IP Core for Configuration of Networking Hardware IPs in the Smart Grid

    Directory of Open Access Journals (Sweden)

    Marcelo Urbina

    2018-02-01

    Full Text Available Nowadays, the incorporation and constant evolution of communication networks in the electricity sector have given rise to the so-called Smart Grid, which is why it is necessary to have devices that are capable of managing new communication protocols, guaranteeing the strict requirements of processing required by the electricity sector. In this context, intelligent electronic devices (IEDs with network architectures are currently available to meet the communication, real-time processing and interoperability requirements of the Smart Grid. The new generation IEDs include an Field Programmable Gate Array (FPGA, to support specialized networking switching architectures for the electric sector, as the IEEE 1588-aware High-availability Seamless Redundancy/Parallel Redundancy Protocol (HSR/PRP. Another advantage to using an FPGA is the ability to update or reconfigure the design to support new requirements that are being raised to the standards (IEC 61850. The update of the architecture implemented in the FPGA can be done remotely, but it is necessary to establish a cyber security mechanism since the communication link generates vulnerability in the case the attacker gains physical access to the network. The research presented in this paper proposes a secure protocol and Intellectual Property (IP core for configuring and monitoring the networking IPs implemented in a Field Programmable Gate Array (FPGA. The FPGA based implementation proposed overcomes this issue using a light Layer-2 protocol fully implemented on hardware and protected by strong cryptographic algorithms (AES-GCM, defined in the IEC 61850-90-5 standard. The proposed secure protocol and IP core are applicable in any field where remote configuration over Ethernet is required for IP cores in FPGAs. In this paper, the proposal is validated in communications hardware for Smart Grids.

  12. Reverse depletion method for PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.; Kim, Y.J.

    1985-01-01

    Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. Prospects for even greater use of in-board fresh fuel loading are good as utilities seek to reduce core vessel fluence, mitigate pressurized thermal shock concerns, and extend vessel lifetime. Consequently, large numbers of burnable poison (BP) pins are being used to control the power peaking at the in-board fresh fuel positions. This has presented an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of BP pins in the fresh fuel. A procedure was developed that utilizes the well-known Haling depletion to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and BP distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provide for the maximum utility to PWR core reload design

  13. Design Of Cooling Configuration For Military Aeroengine V-Gutter

    Directory of Open Access Journals (Sweden)

    Batchu Suresh

    2017-07-01

    Full Text Available Military aircraft engines employ afterburner system for increasing the thrust required during combat and take-off flight conditions. V-gutter is employed for stabilisation of the flame during reheat. For fifth generation aero engine the gas temperature at the start of the afterburner is be-yond the allowable material limits of the V-gutter so it is required to cool the V-gutter to obtain acceptable creep life. The design of cooling configuration for the given source pressure is worked out for different rib configurations to obtain the allowable metal temperature with minimum coolant mass flow.1D network analysis is used to estimate the cooling mass flow and metal temperature for design flight condition. CFD analysis is carried out for four cooling configurations with different rib orientations. Out of four configurations one configuration is selected for the best cooling configuration.

  14. Design of a holographic waveguide with L configuration

    Science.gov (United States)

    Xiang, Guangxin-Xin; Li, Wen-Qiang

    2016-10-01

    In order to decrease the complexity to design and manufacture the turning grating of the configuration with one reflecting surface, an L-shape two-dimension extended configuration with single plate is given in the paper. This configuration consists of one specular reflecting surface and three holographic gratings two in which periods and the groove orientations are totally same, which makes gratings design and fabrication easier. According to the calculation and analysis to the optical path of configuration, the dimension of the turning grating is no larger than 40mm×30mm. The simulation result demonstrates the display configuration is reasonable and correct and can realize the display effect with 30°×30° field of view and Φ30mm large exit pupil. This configuration can be applied to an Augmented Reality Display (AR) or a Head-Mounted Display (HMD).

  15. Designing for Reuse of Configurable Logic

    National Research Council Canada - National Science Library

    Elm, Joseph P

    2005-01-01

    Field-programmable gate arrays (FPGAs) offer electronic systems designers the opportunity to reduce development cost, reduce time-to-market, increase system performance, and improve system adaptability...

  16. Building configuration and seismic design: The architecture of earthquake resistance

    Science.gov (United States)

    Arnold, C.; Reitherman, R.; Whitaker, D.

    1981-05-01

    The architecture of a building in relation to its ability to withstand earthquakes was determined. Aspects of round motion which are significant to building behavior are discussed. Results of a survey of configuration decisions that affect the performance of buildings with a focus on the architectural aspects of configuration design are provided. Configuration derivation, building type as it relates to seismic design, and seismic design, and seismic issues in the design process are examined. Case studies of the Veterans' Administration Hospital in Loma Linda, California, and the Imperial Hotel in Tokyo, Japan, are presented. The seismic design process is described paying special attention to the configuration issues. The need is stressed for guidelines, codes, and regulations to ensure design solutions that respect and balance the full range of architectural, engineering, and material influences on seismic hazards.

  17. Configuration management of the EU DEMO conceptual design data

    International Nuclear Information System (INIS)

    Meszaros, Botond; Shannon, Mark; Marzullo, Domenico; Woodley, Colin; Rowe, Steve; Di Gironimo, Giuseppe

    2016-01-01

    Highlights: • Description of the selection of the DEMO Product Data Management tool. • Introduction of the DEMO configuration management philosophy for the CAD design data. • Description of the enabling tools and systems of the configuration management. - Abstract: The EUROfusion Consortium is setting up – as part of the EU Fusion Roadmap – the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems “light weight” and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes.

  18. Configuration management of the EU DEMO conceptual design data

    Energy Technology Data Exchange (ETDEWEB)

    Meszaros, Botond; Shannon, Mark [EUROfusion Consortium, PPPT Department, Garching, Boltzmannstr. 2 (Germany); Marzullo, Domenico [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Woodley, Colin; Rowe, Steve [CCFE, Culham Science Centre, Oxfordshire OX14 3DB, Abingdon (United Kingdom); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy)

    2016-11-01

    Highlights: • Description of the selection of the DEMO Product Data Management tool. • Introduction of the DEMO configuration management philosophy for the CAD design data. • Description of the enabling tools and systems of the configuration management. - Abstract: The EUROfusion Consortium is setting up – as part of the EU Fusion Roadmap – the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems “light weight” and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes.

  19. Research Design – Composition, Configuration and Interdependencies

    DEFF Research Database (Denmark)

    Kastberg, Peter

    2016-01-01

    In this essay I will present an integrative view on research design. I will introduce what I take to be the skeleton components of any research design within the social sciences, i.e. the elements of research question, philosophy of science, methodology, method and data. With this as my point...... of departure I will go on to focus on a presentation, a discussion and an evaluation of a new appreciation of the interdependencies of the elements in the research design. An appreciation that favors a relational rather than an atomistic outlook and which gives rise to an ecological conceptualization......). At a more concrete level, it aims at presenting an idea of research design which would – hopefully – be an inspiration to (young) scholars....

  20. A review of the core catcher design in LMR

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Hahn, Do Hee

    2001-08-01

    The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, The core catcher design technologies and processes are presented. Finally the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized and the preliminary assessment on the core catcher installation in KALIMER is presented

  1. Configuration Design of Detector Shielding for Gamma Prompt Analysis

    International Nuclear Information System (INIS)

    Elin-Nuraini; Darsono; Elisabeth

    2000-01-01

    Configuration on design of detector shielding for gamma prompt analysishas been performed. The aim of this design is to obtain effective shieldingmaterial and configuration that able to protect the detector for fastneutron. The result shown that detector shielding configuration that obtainedby configuration of water and concrete, would be able to absorb fast neutronup to 99.5 %. The neutron flux that passed through shielding configuration is2.4 x 10 3 n/cm 2 dt, in the detector position of 60 cm (forward neutron beamdirection) on the X axis and 30 cm (side ward neutron beam direction) on theZ axis of target. On this position (60,30) counting result was 104358 for Pbcollimator and 246652 for PVC collimator. From examination result shown thatthe weight of silicon is in order 175 gram. (author)

  2. Development of core design technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; In, Kim Young; Kim, Young Il; Kim, Y G; Kim, S J; Song, H; Kim, T K; Kim, W S; Hwang, W; Lee, B O; Park, C K; Joo, H K; Yoo, J W; Kang, H Y; Park, W S

    2000-05-01

    For the development of KALIMER (150 MWe) core conceptual design, design evolution and optimization for improved economics and safety enhancement was performed in the uranium metallic fueled equilibrium core design which uses U-Zr binary fuel not in excess of 20 percent enrichment. Utilizing results of the uranium ,metallic fueled core design, the breeder equilibrium core design with breeding ratio being over 1.1 was developed. In addition, utilizing LMR's excellent neutron economy, various core concepts for minor actinide burnup, inherent safety, economics and non-proliferation were realized and its optimization studies were performed. A code system for the LMR core conceptual design has been established through the implementation of needed functions into the existing codes and development of codes. To improve the accuracy of the core design, a multi-dimensional nodal transport code SOLTRAN, a three-dimensional transient code analysis code STEP, MATRA-LMR and ASSY-P for T/H analysis are under development. Through the automation of design calculations for efficient core design, an input generator and several interface codes have been developed. (author)

  3. Development of core design technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim Young In; Kim, Young Il; Kim, Y. G.; Kim, S. J.; Song, H.; Kim, T. K.; Kim, W. S.; Hwang, W.; Lee, B. O.; Park, C. K.; Joo, H. K.; Yoo, J. W.; Kang, H. Y.; Park, W. S

    2000-05-01

    For the development of KALIMER (150 MWe) core conceptual design, design evolution and optimization for improved economics and safety enhancement was performed in the uranium metallic fueled equilibrium core design which uses U-Zr binary fuel not in excess of 20 percent enrichment. Utilizing results of the uranium ,metallic fueled core design, the breeder equilibrium core design with breeding ratio being over 1.1 was developed. In addition, utilizing LMR's excellent neutron economy, various core concepts for minor actinide burnup, inherent safety, economics and non-proliferation were realized and its optimization studies were performed. A code system for the LMR core conceptual design has been established through the implementation of needed functions into the existing codes and development of codes. To improve the accuracy of the core design, a multi-dimensional nodal transport code SOLTRAN, a three-dimensional transient code analysis code STEP, MATRA-LMR and ASSY-P for T/H analysis are under development. Through the automation of design calculations for efficient core design, an input generator and several interface codes have been developed. (author)

  4. Analysis of Gamma Dose Rate for RTP 2 MW Core Configuration Using MCNP

    International Nuclear Information System (INIS)

    Mohamad Hairie Rabir; Mohd Amin Sharifuldin Salleh; Julia Abdul Karim

    2011-01-01

    The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the calculation of gamma dose rate at water pool surface and concrete shielding surface of the proposed 2-MW core configuration of PUSPATI TRIGA Reactor. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core with pool water and concrete shielding and validation of the input by comparisons with the measured and available safety analysis report (SAR) of the reactor. The model represents in detailed all components of the reactor with literally no physical approximation. Continuous energy cross section data from the more recent nuclear data as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed. (author)

  5. Exergy analysis of an IGCC design configuration for Plant Wansley

    International Nuclear Information System (INIS)

    Tsatsaronis, G.; Tawfik, T.; Lin, L.; Gallaspy, D.T.

    1989-01-01

    An integrated gasification-combined-cycle power plant design was developed for Georgia Power Company's Plant Wansley. This paper discusses the plant configuration and presents the most important results obtained from a detailed exergy analysis of the plant design. This analysis will be completed in a subsequent paper through an exergoeconomic analysis to identify design improvements for reducing the electricity cost

  6. Web-based Core Design System Development

    International Nuclear Information System (INIS)

    Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

    2011-01-01

    The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

  7. Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

    International Nuclear Information System (INIS)

    Shin, Hyeong-Ki

    1999-01-01

    The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

  8. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs

  9. Preconceptual ABC design definition and system configuration layout: Appendix A

    International Nuclear Information System (INIS)

    1995-03-01

    The mission of the ABC system is to destroy as effectively as possible the fissile material inserted into the core without producing any new fissile material. The contents of this report are as follows: operating conditions for the steam-cycle ABC system; flow rates and component dimensions; drawings of the ABC layout; and impact of core design parameters on containment size

  10. Configurable intelligent optimization algorithm design and practice in manufacturing

    CERN Document Server

    Tao, Fei; Laili, Yuanjun

    2014-01-01

    Presenting the concept and design and implementation of configurable intelligent optimization algorithms in manufacturing systems, this book provides a new configuration method to optimize manufacturing processes. It provides a comprehensive elaboration of basic intelligent optimization algorithms, and demonstrates how their improvement, hybridization and parallelization can be applied to manufacturing. Furthermore, various applications of these intelligent optimization algorithms are exemplified in detail, chapter by chapter. The intelligent optimization algorithm is not just a single algorit

  11. Design study on metal fuel FBR cores

    International Nuclear Information System (INIS)

    Yokoo, T.; Tanaka, Y.; Ogata, T.

    1991-01-01

    A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

  12. Preliminary core design calculations for the ACPR Upgrade

    International Nuclear Information System (INIS)

    Pickard, P.S.

    1976-01-01

    The goal of the Annular Core Pulse Reactor (ACPR) Upgrade design studies is to define a core configuration that provides a significant increase in pulse fluence and fission energy deposition. The reactor modification should provide as flat an energy deposition profile for experiments as feasible. The fuels examined in this study were UO 2 -BeO (5-15 w/o UO 2 ), UC-ZrC-C (200-500 mg U/cc) and U-ZrH 1.5 . The basic core concept examined was a two region core, - a high heat capacity inner core region surrounded by an outer U-ZrH 1.5 region. Survey core calculations utilizing 1D transport calculations and cross sections libraries derived from the ORNL-AMPX code examined relative fuel loadings, fuel temperatures, reactivity requirements and pulse performance improvement. Reference designs for all candidate fuels were defined utilizing 2D transport and Monte Carlo calculations. The performance implications of alternative core designs were also examined for the UO 2 -BeO and UC-ZrC-C fuel candidates. (author)

  13. Preconceptual ABC design definition and system configuration layout

    International Nuclear Information System (INIS)

    Barthold, W.

    1995-03-01

    This document is the conceptual design document for the follow-on to the Molten Salt Breeder Reactor, known as the ABC type reactor. It addresses blanket design options, containment options, off-gas systems, drainage systems, and components/layouts of the primary, secondary, and tertiary systems, and it contains a number of diagrams for the configuration of the major systems

  14. Design of full MOX core in ABWR

    International Nuclear Information System (INIS)

    Kinoshita, Y.; Hirose, T.; Sasagawa, M.; Sakuma, T

    1999-01-01

    A Full MOX-ABWR, loaded with mixed-oxide (MOX) fuels of up to 100% of the core, is planned. Increased MOX fuel utilization will result in greater savings of uranium. Studies on the fuel rod thermal-mechanical design, the core design and the safety evaluation have been made, and the results are summarized in this paper. To sum it all up, the safety of the Full MOX-ABWR has been confirmed through design evaluations adequately considering the MOX fuel and core characteristics. (author)

  15. Core design and fuel management studies

    International Nuclear Information System (INIS)

    Min, Byung Joo; Chan, P.

    1997-06-01

    The design target for the CANDU 9 requires a 20% increase in electrical power output from an existing 480-channel CANDU core. Assuming a net electrical output of 861 MW(e) for a natural uranium fuelled Bruce-B/Darlington reactor in a warm water site, the net electrical output of the reference CANDU 9 reactor would be 1033 MW(e). This report documents the result of the physics studies for the design of the CANDU 9 480/SEU core. The results of the core design and fuel management studies of the CANDU 9 480/SEU reactor indicated that up to 1033 MW(e) output can be achieved in a 480-channel CANDU core by using SEU core can easily be maintained indefinitely using an automated refuelling program. Fuel performance evaluation based on the data of the 500 FPDs refuelling simulation concluded that SEU fuel failure is not expected. (author). 2 tabs., 38 figs., 5 refs

  16. A Design of Product Collaborative Online Configuration Model

    Science.gov (United States)

    Wang, Xiaoguo; Zheng, Jin; Zeng, Qian

    According to the actual needs of mass customization, the personalization of product and its collaborative design, the paper analyzes and studies the working mechanism of modular-based product configuration technology and puts forward an information model of modular product family. Combined with case-based reasoning techniques (CBR) and the constraint satisfaction problem solving techniques (CSP), we design and study the algorithm for product configuration, and analyze its time complexity. A car chassis is made as the application object, we provide a prototype system of online configuration. Taking advantage of this system, designers can make appropriate changes on the existing programs in accordance with the demand. This will accelerate all aspects of product development and shorten the product cycle. Also the system will provide a strong technical support for enterprises to improve their market competitiveness.

  17. Nuclear design and analysis report for KALIMER breakeven core conceptual design

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

    2002-04-01

    During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

  18. District Heating Network Design and Configuration Optimization with Genetic Algorithm

    DEFF Research Database (Denmark)

    Li, Hongwei; Svendsen, Svend

    2013-01-01

    In this paper, the configuration of a district heating network which connects from the heating plant to the end users is optimized. Each end user in the network represents a building block. The connections between the heat generation plant and the end users are represented with mixed integer...... and the pipe friction and heat loss formulations are non-linear. In order to find the optimal district heating network configuration, genetic algorithm which handles the mixed integer nonlinear programming problem is chosen. The network configuration is represented with binary and integer encoding...... and it is optimized in terms of the net present cost. The optimization results indicates that the optimal DH network configuration is determined by multiple factors such as the consumer heating load, the distance between the heating plant to the consumer, the design criteria regarding the pressure and temperature...

  19. Reactor core design aiding system

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Hamaguchi, Yukio; Nakao, Takashi; Kondo, Yasuhide

    1995-01-01

    A two-dimensional radial power distribution and an axial one-dimensional power distribution are determined based on a distribution of a three-dimensional infinite multiplication factor, to obtain estimated power distribution estimation values. The estimation values are synthesized to obtain estimated three-dimensional power distribution values. In addition, the distribution of a two-dimensional radial multiplication factor and the distribution of an one-dimensional axial multiplication factor are determined based on the three-dimensional power distribution, to obtain estimated values for the multiplication factor distribution. The estimated values are synthesized to form estimated values for the three-dimensional multiplication factor distribution. Further, estimated fuel loading pattern value is determined based on the three-dimensional power distribution or the two-dimensional radial power distribution. Since the processes for determining the estimated values comprise only additive and multiplying operations, processing time can be remarkably saved compared with calculation based on a detailed physical models. Since the estimation is performed on every fuel assemblies, a nervous circuit network not depending on the reactor core system can be constituted. (N.H.)

  20. Preliminary core design of IRIS-50

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Franceschini, Fausto

    2009-01-01

    IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

  1. Status of a reformer design for a modular HTGR in an in-line configuration

    International Nuclear Information System (INIS)

    Gluck, R.; Whitling, W.H.; Lipps, A.J.

    1984-01-01

    For the past several years the General Electric Company has had the technical lead on advanced concept studies for the Modular High Temperature Gas Cooled Reactor (HTGR) programs sponsored by the United States Department of Energy. The focus of the Modular Reactor System (MRS) effort is the development of a generic nuclear heat source capable of supplying heat to either a steam generator/electric cycle or a high temperature steam /methane reforming cycle. Some early ground rules for this study were that the reactor be designed for 950 deg. C direct cycle reforming and that the core be of the prismatic type. Since the prismatic core required control rods near the center of the core, the vertical in-line concept was selected to promote natural circulation cooling of the core for all potential transients except the depressurized core heatup transient. Although the requirement for a prismatic core has been eliminated for recent cost reduction studies, the vertical in-line configuration has been retained for its potential as the lowest cost configuration. This paper presents the results of recent design and analytical studies conducted to evaluate the feasibility of using a steam/methane reformer in a Vertical In-Line (VIL) arrangement with the generic nuclear heat source

  2. Development of Core Design Technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

    2007-06-15

    This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

  3. CORD, PWR Core Design and Fuel Management

    International Nuclear Information System (INIS)

    Trkov, Andrej

    1996-01-01

    1 - Description of program or function: CORD-2 is intended for core design applications of pressurised water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refuelling). 2 - Method of solution: The calculations are performed at the cell level with a lattice code in the supercell approximation to generate the single cell cross sections. Fuel assembly cross section homogenization is done in the diffusion approximation. Global core calculations can be done in the full three-dimensional cartesian geometry. Thermohydraulic feedbacks can be accounted for. The Effective Diffusion Homogenization method is used for generating the homogenized cross sections. 3 - Restrictions on the complexity of the problem: The complexity of the problem is selected by the user, depending on the capacity of his computer

  4. Natural convection of the oxide pool in a three-layer configuration of core melts

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Su-Hyeon; Park, Hae-Kyun; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

    2017-06-15

    Highlights: • Natural convection of oxide pool in 3-layer configuration during IVR was investigated. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Heat ratio to light metal layer was 14% higher for 3-layer configuration than 2-layer one. • Heat transfer to heavy metal layer was poor and hence heat load to side wall increased. • Angular heat loads to side wall showed strengthened heat focusing at uppermost location. - Abstract: We investigated the natural convection of the oxide layer in a three-layer configuration of core melts in a severe accident. In order to achieve high modified Rayleigh numbers of 10{sup 12}–10{sup 13}, mass transfer experiments were performed using a copper sulfate electroplating system based upon the analogy between heat and mass transfer. Four different cooling conditions of the top and the bottom plates were tested. The upward heat ratios were 14% higher for three-layer than for two-layer due to the reduced heights and the downward heat ratios were lower the same amount. The local Nusselt numbers for the top and the bottom plates were measured and compared with the two layer configuration. To explore the heat load to the reactor vessel, the angle-dependent heat fluxes at the side wall, were measured and compared with the two-layer configuration. Heat load to the side wall and peak heat at the uppermost location were intensified for the three-layer configuration.

  5. Design basis reconstitution and configuration management of nuclear power plants

    International Nuclear Information System (INIS)

    Smith, P.R.

    1989-01-01

    The major design requirements of nuclear power plant components, systems, and structures are found in the plant's licensing commitments documented in the Final Safety Analysis Report and in the technical specification commitments of the plant. These specifications consider the original design and its degradation by in-service use. Before a nuclear power plant begins operation, the plant systems, structures, and organizational elements are functionally arranged to operate in a particular way. This functional arrangement is specified by the plant's design requirements and is called its configuration. The paper discusses configuration management and information management for configuration management. The management of large amounts of information and the various information systems associated with nuclear generating facilities is an ever-growing challenge for utilities. Plant operations involve a complex interrelation among data elements, especially in relation to design modifications and operational changes. Consequently, the operation of these data systems is interrelated and, as a result, redundant data items may exist. Thus, in view of the need to control and manage the plant configuration baseline, managers are striving to streamline their information management programs, which usually involves the integration of data-base systems

  6. Nuclear structure calculations in $^{20}$Ne with No-Core Configuration-Interaction model

    OpenAIRE

    Konieczka, Maciej; Satuła, Wojciech

    2016-01-01

    Negative parity states in $^{20}$Ne and Gamow-Teller strength distribution for the ground-state beta-decay of $^{20}$Na are calculated for the very first time using recently developed No-Core Configuration-Interaction model. The approach is based on multi-reference density functional theory involving isospin and angular-momentum projections. Advantages and shortcomings of the method are briefly discussed.

  7. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Clark, J.S.; Walton, J.T.; Mcguire, M.L.

    1992-07-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

  8. The Study of Spherical Cores with a Toroidal Magnetic Field Configuration

    Energy Technology Data Exchange (ETDEWEB)

    Gholipour, Mahmoud [Research Institute for Astronomy and Astrophysics of Maragha (RIAAM)—Maragha, P.O. Box 55134-441 (Iran, Islamic Republic of)

    2017-04-01

    Observational studies of the magnetic fields in molecular clouds have significantly improved the theoretical models developed for the structure and evolution of dense clouds and for the star formation process as well. The recent observational analyses on some cores indicate that there is a power-law relationship between magnetic field and density in the molecular clouds. In this study, we consider the stability of spherical cores with a toroidal magnetic field configuration in the molecular clouds. For this purpose, we model a spherical core that is in magnetostatic equilibrium. Herein, we propose an equation of density structure, which is a modified form of the isothermal Lane–Emden equation in the presence of the toroidal magnetic field. The proposed equation describes the effect of the toroidal magnetic field on the cloud structure and the mass cloud. Furthermore, we found an upper limit for this configuration of magnetic field in the molecular clouds. Then, the virial theorem is used to consider the cloud evolution leading to an equation in order to obtain the lower limit of the field strength in the molecular cloud. However, the results show that the field strength of the toroidal configuration has an important effect on the cloud structure, whose upper limit is related to the central density and field gradient. The obtained results address some regions of clouds where the cloud decomposition or star formation can be seen.

  9. Development of small, fast reactor core designs using lead-based coolant

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

    1999-01-01

    A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

  10. Core polarisation and configuration mixing in 58Ni studied by high resolution electron scattering

    International Nuclear Information System (INIS)

    Blok, H.

    1986-01-01

    The nucleus 58 Ni is studied by inelastic electron-scattering. This nucleus has two valence neutrons outside a closed 58 Ni core which implies that no valence protons contribute to the transitions and thus, besides configuration mixing of the valence neutrons, proton-core polarization can be studied in detail. From inelastic electron-scattering data one obtains the charge- and current-transition densities by determining the Fourier-Bessel transform of the cross sections measured over a wide range of linear momenta transferred to the nucleus. The results of an analysis of the excitation of two 0 ++ states at low-momentum transfer are presented. These transitions are particularly interesting for studying core-polarization contributions. (Auth.)

  11. Multiple spacecraft configuration designs for coordinated flight missions

    Science.gov (United States)

    Fumenti, Federico; Theil, Stephan

    2018-06-01

    Coordinated flight allows the replacement of a single monolithic spacecraft with multiple smaller ones, based on the principle of distributed systems. According to the mission objectives and to ensure a safe relative motion, constraints on the relative distances need to be satisfied. Initially, differential perturbations are limited by proper orbit design. Then, the induced differential drifts can be properly handled through corrective maneuvers. In this work, several designs are surveyed, defining the initial configuration of a group of spacecraft while counteracting the differential perturbations. For each of the investigated designs, focus is placed upon the number of deployable spacecraft and on the possibility to ensure safe relative motion through station keeping of the initial configuration, with particular attention to the required Δ V budget and the constraints violations.

  12. LMFBR design and its evolution. (2) Core design of LMFBR

    International Nuclear Information System (INIS)

    Uto, Nariaki; Mizuno, Tomoyasu

    2003-01-01

    Sodium-cooled core design studies are performed. MOX fuel core with axial blanket partial elimination subassembly due to safety consideration is studied. This type of core with high internal conversion ratio possesses capability of achieving 26 months of operation cycle length and 100 GWd/t of burnup averaged over core and blanket, which are superior characteristics in view of reducing cost of power generation. Metal fuel core is also studied, and its higher breeding capability reveals a potential of better core performance such as longer operation cycle length for the same level of electricity generation, though core outlet temperature is limited to lower level due to steel cladding-metal fuel compatibility concerns. Another metal fuel core concept using single Pu enrichment and two radial regions with individual fuel pin diameters achieves 550degC of core outlet temperature identical to that of MOX fuel core, keeping operation cycle length comparable with that of MOX fuel core. This series of study results show that sodium-cooled MOX and metal fuel cores have a high flexibility in satisfying various needs including fuel cycle cost and breeding capability, depending on the stage of introducing commercialized fast reactor cycle system. (author)

  13. Innovative reactor core: potentialities and design

    International Nuclear Information System (INIS)

    Artioli, C.; Petrovich, Carlo; Grasso, Giacomo

    2010-01-01

    Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GW e y. (authors)

  14. Full MOX core design in ABWR

    International Nuclear Information System (INIS)

    Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

    2003-01-01

    Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

  15. Conceptual Design of the RHIC Dump Core

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, A. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    1995-09-26

    Conceptually, the internal dump consists of a "core" whose purpose is to absorb the energy of the beam, and surrounding shielding whose purpose is to attenuate radiation. Design of the core for an internal dump has two problems which must be overcome. The first problem is preserving the integrity of the dump core. The bunches must be dispersed laterally an amount sufficient to keep the energy density from cracking the dump core material. Since the dump kickers in RHIC are only ~25m upstream of the entrance face of the dump, this is i a difficult problem. The second problem, not addressed in this note, is that dumping the beam should not quench downstream magnets. Preliminary calculations related to both of these problems have been presented in earlier notes.

  16. Fuel management strategy for the compact core design of RSG GAS (MPR-30)

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Liem, P.H.; Tukiran, S. [National Nuclear Energy Agency (Batan), PUSPIPTEK-Serpong Tangerang (Indonesia)

    2000-07-01

    The rearrangement of the core configuration of the RSG GAS reactor to obtain a compact core is in progress. A fuel management strategy is proposed for the equilibrium compact core of this reactor by reducing the number of in-core irradiation positions. The reduced irradiation positions are based on the activities during 12 years operation. The obtained compact core gives significant extension of the operation cycle length so that the reactor availability and utilization can be enhanced. The equilibrium compact silicide core obtained met the imposed design constraints and safety requirements. (author)

  17. Fuel management strategy for the compact core design of RSG GAS (MPR-30)

    International Nuclear Information System (INIS)

    Sembiring, T.M.; Liem, P.H.; Tukiran, S.

    2000-01-01

    The rearrangement of the core configuration of the RSG GAS reactor to obtain a compact core is in progress. A fuel management strategy is proposed for the equilibrium compact core of this reactor by reducing the number of in-core irradiation positions. The reduced irradiation positions are based on the activities during 12 years operation. The obtained compact core gives significant extension of the operation cycle length so that the reactor availability and utilization can be enhanced. The equilibrium compact silicide core obtained met the imposed design constraints and safety requirements. (author)

  18. Some concept for the TRIGA core design

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1994-01-01

    There is the research reactor called TRIGA Mark-2 of 100 kW in Atomic Energy Research Laboratory, Musashi Institute of Technology. Recently, while the various calculations on the core were carried out, the author became aware of that this TRIGA core was designed at that time with excellent consideration. The reason for that is, although fuel is arranged in simple concentric circular state at a glance, it was known that in reality, this is the modification of the hexagonal core of triangular lattice. In the examination of square lattice fuel arrangement, the reactivity was calculated by using the gap between fuel rods as the parameter and by using ENDF/B-4 library and Monte Carlo code Keno-5. It is known that the design of the lattice with maximum reactivity cannot be done by the square lattice. The similar examination was carried out on triangular lattice, and it was found that the gap between fuel rods of 4 mm is the optimal design. The average neutron energy spectra in the fuel rods of the TRIGA Mark-2 core agreed considerably well with the energy spectra at 4.16 cm fuel rod pitch in triangular hexagonal core. In the reactor of about 100 kW, even if the gap between fuel rods is less than 4 mm, heat removal is sufficiently possible. (K.I.)

  19. The reactor core configuration and important systems related to physics tests of Daya Bay NPP

    International Nuclear Information System (INIS)

    Tao Shaoping

    1995-06-01

    A brief introduction to reactor core configuration and important systems related to physics tests of Daya Bay NPP is given. These systems involve the reactor core system (COR), the full length rod control system (RGL), the in-core instrumentation system (RIC), the out-of-core nuclear instrumentation system (RPN), and the LOCA surveillance system (LSS), the centralized data processing system (KIT) and the test data acquisition system (KDO). In addition, that the adjustment and evaluation of boron concentration related to other systems, for example the reactor coolant system (RCP), the chemical and volume control system (RCV), the reactor boron and water makeup system (REA), the nuclear sampling system (REN) and the reactor control system (RRC), etc. is also described. Analysis of these systems helps not only to familiarize their functions and acquires a deepen understanding for the principle procedure, points for attention and technical key of the core physics tests, but also to further analyze the test results. (3 refs., 11 figs., 1 tab.)

  20. Advance of core design method for ATR

    International Nuclear Information System (INIS)

    Maeda, Seiichirou; Ihara, Toshiteru; Iijima, Takashi; Seino, Hideaki; Kobayashi, Tetsurou; Takeuchi, Michio; Sugawara, Satoru; Matsumoto, Mitsuo.

    1995-01-01

    Core characteristics of ATR demonstration plant has been revised such as increasing the fuel burnup and the channel power, which is achieved by changing the number of fuel rod per fuel assembly from 28 to 36. The research and development concerning the core design method for ATR have been continued. The calculational errors of core analysis code have been evaluated using the operational data of FUGEN and the full scale simulated test results in DCA (Deuterium Critical Assembly) and HTL (Heat Transfer Loop) at O-arai engineering center. It is confirmed that the calculational error of power distribution is smaller than the design value of ATR demonstration plant. Critical heat flux correlation curve for 36 fuel rod cluster has been developed and the probability evaluation method based on its curve, which is more rational to evaluate the fuel dryout, has been adopted. (author)

  1. Configuration System for Simulation Based Design of Vibratory Bowl Feeders

    DEFF Research Database (Denmark)

    Hansson, Michael Natapon; Mathiesen, Simon; Ellekilde, Lars-Peter

    2017-01-01

    Vibratory bowl feeders are still among the most commonly used production equipment for automated part feeding, where parts are correctly oriented for further manipulation by being conveyed through a set of orienting devices. Designing vibratory bowl feeders involves selecting and sequencing...... a number of these devices that either reorients or rejects the part until a desirable orientation is achieved. To aid the designer in this task, this work presents a configuration system where knowledge of the behaviour for each device is acquired through dynamic simulation, and used to solve...

  2. The SSC superconducting air core toroid design development

    International Nuclear Information System (INIS)

    Fields, T.; Carroll, A.; Chiang, I.H.; Frank, J.S.; Haggerty, J.; Littenberg, L.; Morse, W.; Strand, R.C.; Lau, K.; Weinstein, R.; McNeil, R.; Friedman, J.; Hafen, E.; Haridas, P.; Kendall, H.W.; Osborne, L.; Pless, I.; Rosenson, L.; Pope, B.; Jones, L.W.; Luton, J.N.; Bonanos, P.; Marx, M.; Pusateri, J.A.; Favale, A.; Gottesman, S.; Schneid, E.; Verdier, R.

    1990-01-01

    Superconducting air core toroids show great promise for use in a muon spectrometer for the SSC. Early studies by SUNY at Stony Brook funded by SSC Laboratory, have established the feasibility of building magnets of the required size. The toroid spectrometer consists of a central toroid with two end cap toroids. The configuration under development provides for muon trajectory measurement outside the magnetic volume. System level studies on support structure, assembly, cryogenic material selection, and power are performed. Resulting selected optimal design and assembly is described. 4 refs., 6 figs

  3. High-Fidelity Multidisciplinary Design Optimization of Aircraft Configurations

    Science.gov (United States)

    Martins, Joaquim R. R. A.; Kenway, Gaetan K. W.; Burdette, David; Jonsson, Eirikur; Kennedy, Graeme J.

    2017-01-01

    To evaluate new airframe technologies we need design tools based on high-fidelity models that consider multidisciplinary interactions early in the design process. The overarching goal of this NRA is to develop tools that enable high-fidelity multidisciplinary design optimization of aircraft configurations, and to apply these tools to the design of high aspect ratio flexible wings. We develop a geometry engine that is capable of quickly generating conventional and unconventional aircraft configurations including the internal structure. This geometry engine features adjoint derivative computation for efficient gradient-based optimization. We also added overset capability to a computational fluid dynamics solver, complete with an adjoint implementation and semiautomatic mesh generation. We also developed an approach to constraining buffet and started the development of an approach for constraining utter. On the applications side, we developed a new common high-fidelity model for aeroelastic studies of high aspect ratio wings. We performed optimal design trade-o s between fuel burn and aircraft weight for metal, conventional composite, and carbon nanotube composite wings. We also assessed a continuous morphing trailing edge technology applied to high aspect ratio wings. This research resulted in the publication of 26 manuscripts so far, and the developed methodologies were used in two other NRAs. 1

  4. Impact of correlations between core configurations for the evaluation of nuclear data uncertainty propagation for reactivity

    International Nuclear Information System (INIS)

    Frosio, T.; Bonaccorsi, T.; Blaise, P.

    2017-01-01

    The precise estimation of Pearson correlations, also called 'representativity' coefficients, between core configurations is a fundamental quantity for properly assessing the nuclear data (ND) uncertainties propagation on integral parameters such as k-eff, power distributions, or reactivity coefficients. In this paper, a traditional adjoint method is used to propagate ND uncertainty on reactivity and reactivity coefficients and estimate correlations between different states of the core. We show that neglecting those correlations induces a loss of information in the final uncertainty. We also show that using approximate values of Pearson does not lead to an important error of the model. This calculation is made for reactivity at the beginning of life and can be extended to other parameters during depletion calculations. (authors)

  5. District Heating Network Design and Configuration Optimization with Genetic Algorithm

    DEFF Research Database (Denmark)

    Li, Hongwei; Svendsen, Svend

    2011-01-01

    In this paper, the configuration of a district heating (DH) network which connects from the heating plant to the end users was optimized with emphasizing the network thermal performance. Each end user in the network represents a building block. The locations of the building blocks are fixed while...... the heating plant location is allowed to vary. The connection between the heat generation plant and the end users can be represented with mixed integer and the pipe friction and heat loss formulations are non-linear. In order to find the optimal DH distribution pipeline configuration, the genetic algorithm...... by multi factors as the consumer heating load, the distance between the heating plant to the consumer, the design criteria regarding pressure and temperature limitation, as well as the corresponding network heat loss....

  6. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  7. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    International Nuclear Information System (INIS)

    Papukchiev, Angel; Schaefer, Anselm

    2008-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  8. Process of establishing design requirements and selecting alternative configurations for conceptual design of a VLA

    Directory of Open Access Journals (Sweden)

    Bo-Young Bae

    2017-04-01

    Full Text Available In this study, a process for establishing design requirements and selecting alternative configurations for the conceptual phase of aircraft design has been proposed. The proposed process uses system-engineering-based requirement-analysis techniques such as objective tree, analytic hierarchy process, and quality function deployment to establish logical and quantitative standards. Moreover, in order to perform a logical selection of alternative aircraft configurations, it uses advanced decision-making methods such as morphological matrix and technique for order preference by similarity to the ideal solution. In addition, a preliminary sizing tool has been developed to check the feasibility of the established performance requirements and to evaluate the flight performance of the selected configurations. The present process has been applied for a two-seater very light aircraft (VLA, resulting in a set of tentative design requirements and two families of VLA configurations: a high-wing configuration and a low-wing configuration. The resulting set of design requirements consists of three categories: customer requirements, certification requirements, and performance requirements. The performance requirements include two mission requirements for the flight range and the endurance by reflecting the customer requirements. The flight performances of the two configuration families were evaluated using the sizing tool developed and the low-wing configuration with conventional tails was selected as the best baseline configuration for the VLA.

  9. Conceptual Models Core to Good Design

    CERN Document Server

    Johnson, Jeff

    2011-01-01

    People make use of software applications in their activities, applying them as tools in carrying out tasks. That this use should be good for people--easy, effective, efficient, and enjoyable--is a principal goal of design. In this book, we present the notion of Conceptual Models, and argue that Conceptual Models are core to achieving good design. From years of helping companies create software applications, we have come to believe that building applications without Conceptual Models is just asking for designs that will be confusing and difficult to learn, remember, and use. We show how Concept

  10. Influence of magnetic field configuration on magnetohydrodynamic waves in Earth's core

    Science.gov (United States)

    Knezek, Nicholas; Buffett, Bruce

    2018-04-01

    We develop a numerical model to study magnetohydrodynamic waves in a thin layer of stratified fluid near the surface of Earth's core. Past studies have been limited to using simple background magnetic field configurations. However, the choice of field distribution can dramatically affect the structure and frequency of the waves. To permit a more general treatment of background magnetic field and layer stratification, we combine finite volume and Fourier methods to describe the wave motions. We validate our model by comparisons to previous studies and examine the influence of background magnetic field configuration on two types of magnetohydrodynamic waves. We show that the structure of zonal Magnetic-Archimedes-Coriolis (MAC) waves for a dipole background field is unstable to small perturbations of the field strength in the equatorial region. Modifications to the wave structures are computed for a range of field configurations. In addition, we show that non-zonal MAC waves are trapped near the equator for realistic magnetic field distributions, and that their latitudinal extent depends upon the distribution of magnetic field strength at the CMB.

  11. On-line generation of core monitoring power distribution in the SCOMS couppled with core design code

    International Nuclear Information System (INIS)

    Lee, K. B.; Kim, K. K.; In, W. K.; Ji, S. K.; Jang, M. H.

    2002-01-01

    The paper provides the description of the methodology and main program module of power distribution calculation of SCOMS(SMART COre Monitoring System). The simulation results of the SMART core using the developed SCOMS are included. The planar radial peaking factor(Fxy) is relatively high in SMART core because control banks are inserted to the core at normal operation. If the conventional core monitoring method is adapted to SMART, highly skewed planar radial peaking factor Fxy yields an excessive conservatism and reduces the operation margin. In addition to this, the error of the core monitoring would be enlarged and thus operating margin would be degraded, because it is impossible to precalculate the core monitoring constants for all the control banks configurations taking into account the operation history in the design stage. To get rid of these drawbacks in the conventional power distribution calculation methodology, new methodology to calculate the three dimensional power distribution is developed. Core monitoring constants are calculated with the core design code (MASTER) which is on-line coupled with SCOMS. Three dimensional (3D) power distribution and the several peaking factors are calculated using the in-core detector signals and core monitoring constant provided at real time. Developed methodology is applied to the SMART core and the various core states are simulated. Based on the simulation results, it is founded that the three dimensional peaking factor to calculate the Linear Power Density and the pseudo hot-pin axial power distribution to calculate the Departure Nucleate Boiling Ratio show the more conservative values than those of the best-estimated core design code, and SCOMS adapted developed methodology can secures the more operation margin than the conventional methodology

  12. Building a design community for sustainable homes through configuration and open innovation

    DEFF Research Database (Denmark)

    Thuesen, Christian Langhoff; Jespersen, Kristina Risom

    2009-01-01

    . Furthermore will an idea space collect and rate ideas from the users. Through a combination of technical and user driven innovation the design community will act as a learning tool for the users and producers and thereby facilitate the development of a market for sustainable homes.......This paper presents a development project which aims to create a market place for sustainable homes – around a design community where the uses and producers collectively can develop new energy efficient solutions and thereby reduce the emmisson of CO2. The core functionality of the design community...... is a configurator where the users based on the produceres templates can design their own home at a selected address visualizing and estimating the energy consumption, total cost, CO2 emission etc. All the designs will be collected and rated in a design space creating transparency over the market and technologies...

  13. Design of materials configurations for enhanced phononic and electronic properties

    Science.gov (United States)

    Daraio, Chiara

    The discovery of novel nonlinear dynamic and electronic phenomena is presented for the specific cases of granular materials and carbon nanotubes. This research was conducted for designing and constructing optimized macro-, micro- and nano-scale structural configurations of materials, and for studying their phononic and electronic behavior. Variation of composite arrangements of granular elements with different elastic properties in a linear chain-of-sphere, Y-junction or 3-D configurations led to a variety of novel phononic phenomena and interesting physical properties, which can be potentially useful for security, communications, mechanical and biomedical engineering applications. Mechanical and electronic properties of carbon nanotubes with different atomic arrangements and microstructures were also investigated. Electronic properties of Y-junction configured carbon nanotubes exhibit an exciting transistor switch behavior which is not seen in linear configuration nanotubes. Strongly nonlinear materials were designed and fabricated using novel and innovative concepts. Due to their unique strongly nonlinear and anisotropic nature, novel wave phenomena have been discovered. Specifically, violations of Snell's law were detected and a new mechanism of wave interaction with interfaces between NTPCs (Nonlinear Tunable Phononic Crystals) was established. Polymer-based systems were tested for the first time, and the tunability of the solitary waves speed was demonstrated. New materials with transformed signal propagation speed in the manageable range of 10-100 m/s and signal amplitude typical for audible speech have been developed. The enhancing of the mitigation of solitary and shock waves in 1-D chains were demonstrated and a new protective medium was designed for practical applications. 1-D, 2-D and 3-D strongly nonlinear system have been investigated providing a broad impact on the whole area of strongly nonlinear wave dynamics and creating experimental basis for new

  14. Core design aspects of SNR 2

    International Nuclear Information System (INIS)

    Wehmann, U.K.

    1987-01-01

    The paper describes in its first part the main characteristics of the core of the SNR 2 fast breeder reactor which is being planned within the European collaboration on fast breeder reactors. In the second part some core design aspects are discussed. The fuel element management with an inwards shuffling after each cycle is illustrated which offers advantages with respect to linear rating, steel damage and average discharge burnup. For this management, the full three-dimensional power and burnup history has been calculated and some typical results are presented. The shutdown requirements and the capabilities of the two shutdown systems of SNR 2 are discussed. The necessity for a reliable surveillance of the power distribution is demonstrated by the pronounced power tilts in case of the unintentional withdrawal of an absorber rod. Finally, a short review of the main nuclear design methods and their validation with help of the evaluation of experiments in zero power facilities and power reactors is given

  15. Graphite core design in UK reactors

    International Nuclear Information System (INIS)

    Davies, M.W.

    1996-01-01

    The cores in the first power producing Magnox reactors in the UK were designed with only a limited amount of information available regarding the anisotropic dimensional change behaviour of Pile Grade graphite. As more information was gained it was necessary to make modifications to the design, some minor, some major. As the cores being built became larger, and with the switch to the Advanced Gas-cooled Reactor (AGR) with its much higher power density, additional problems had to be overcome such as increased dimensional change and radiolytic oxidation by the carbon dioxide coolant. For the AGRs a more isotropic graphite was required, with a lower initial open pore volume and higher strength. Gilsocarbon graphite was developed and was selected for all the AGRs built in the UK. Methane bearing coolants are used to limit radiolytic oxidation. (author). 5 figs

  16. Thermal hydraulics and mechanics core design programs

    International Nuclear Information System (INIS)

    Heinecke, J.

    1992-10-01

    The report documents the work performed within the Research and Development Task T hermal hydraulics and mechanics core design programs , funded by the German government. It contains the development of new codes, the extension of existing codes, the qualification and verification of codes and the development of a code library. The overall goal of this work was to adapt the system of thermal hydraulics and mechanics codes to the permanently growing requirements of the status of science and technology

  17. Thermal hydraulic design of PFBR core

    International Nuclear Information System (INIS)

    Roychowdhury, D.G.; Vinayagam, P.P.; Ravichandar, S.C.

    2000-01-01

    The thermal-hydraulic design of core is important in respecting temperature limits while achieving higher outlet temperature. This paper deals with the analytical process developed and implemented for analysing steady state thermal-hydraulics of PFBR core. A computer code FLONE has been developed for optimisation of flow allocation through the subassemblies (SA). By calibrating β n (ratio between the maximum channel temperature rise and SA average temperature rise) values with SUPERENERGY code and using these values in FLONE code, prediction of average and maximum coolant temperature distribution is found to be reasonably accurate. Hence, FLONE code is very powerful design tool for core design. A computer code SAPD has been developed to calculate the pressure drop of fuel and blanket SA. Selection of spacer wire pitch depends on the pressure drop, flow-induced vibration and the mixing characteristics. A parametric study was made for optimisation of spacer wire pitch for the fuel SA. Experimental programme with 19 pin-bundle has been undertaken to find the flow-induced vibration characteristics of fuel SA. Also, experimental programme has been undertaken on a full-scale model to find the pressure drop characteristics in unorificed SA, orifices and the lifting force on the SA. (author)

  18. Design and analysis of EI core structured transverse flux linear reluctance actuator

    OpenAIRE

    FENERCİOĞLU, AHMET; AVŞAR, YUSUF

    2015-01-01

    In this study, an EI core linear actuator is proposed for horizontal movement systems. It is a transverse flux linear switched reluctance motor designed with an EI core structure geometrically. The actuator is configured into three phases and at a 6/4 pole ratio, and it has a stationary active stator along with a sliding passive translator. The stator consists of E cores and the translator consists of I cores. The actuator has a yokeless design because the stator and translator have no back i...

  19. Core designs of modern VVER projects

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Kushmanov, S.; Vjalitsyn, V.; Vasilchenko, R.

    2015-01-01

    The presented operational experience of TVS - 2M (pilot-commercial operation started in 2006 at Balakovo NPP -1) enables to use it as reference for new projects because of similarity in designs and operational conditions. In the paper main parameters of fuel cycles, stability to impact of damaging factors, pilot operation of MG, new alloys, ADF and NTMC, upgrade of FA - 2M for the further power uprating, profiling of Gd-fuel rods for 18-month Fuel Cycle (FC) and perfection of absorber element design are the discussed issues. At the end author concluded that: 1) Core designs of new projects AES-2006 and VVER-TOI are based on extensive successful operational experience of the close prototype of TVS - 2M. 2) All improvements both of technical and economic parameters of fuel are subjected to representative examination by pilot operation at the power units with VVER-1000 being close prototypes of new designs

  20. SMART core preliminary nuclear design-II

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

    1997-06-01

    Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

  1. Computer-Aided Test Flow in Core-Based Design

    OpenAIRE

    Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

    2000-01-01

    This paper copes with the test-pattern generation and fault coverage determination in the core based design. The basic core-test strategy that one has to apply in the core-based design is stated in this work. A Computer-Aided Test (CAT) flow is proposed resulting in accurate fault coverage of embedded cores. The CAT now is applied to a few cores within the Philips Core Test Pilot IC project

  2. Core design characteristics of the hyper system

    International Nuclear Information System (INIS)

    Yonghee, Kim; Won-Seok, Park; Hill, R.N.

    2003-01-01

    In Korea, an accelerator-driven system (ADS) called HYPER (Hybrid Power Extraction Reactor) is being studied for the transmutation of the radioactive wastes. HYPER is a 1000 MWth lead-bismuth eutectic (LBE)-cooled ADS. In this paper, the neutronic design characteristics of HYPER are described and its transmutation performances are assessed for an equilibrium cycle. The core is loaded with a ductless fuel assembly containing transuranics (TRU) dispersion fuel pins. In HYPER, a relatively high core height, 160 cm, is adopted to maximize the multiplication efficiency of the external source. In the ductless fuel assembly, 13 non-fuel rods are used as tie rods to maintain the mechanical integrity of assembly. As the reflector material, pure lead is used to improve the neutron economy and to minimise the generation of radioactive materials. In HYPER, to minimise the burn-up reactivity swing, a B 4 C burnable absorber is employed. For efficient depletion of the B-10 absorber, the burnable absorber is loaded only in the axially-central part (92 cm long) of the 13 tie rods of each assembly. In the current design, the amount of the B 4 C absorber was determined such that the burn-up reactivity swing is about 3.0% Δk. The long-lived fission products (LLFPs) 99 Tc and 129 I are also transmuted in the HYPER core such that their supporting ratios are equal to that of the TRUs. A heterogeneous LLFP transmutation in the reflector zone has been analysed in this work. A unique feature of the HYPER system is that it has an auxiliary core shutdown system, independent of the accelerator shutdown system. It has been shown that a cylindrical B 4 C absorber between the target and fuel blanket can drastically reduce the fission power even without shutting off the accelerator power. (author)

  3. Effect of core configuration on the burnup calculations of MTR research reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Amin, E.H.; Sakr, A.M.

    2014-01-01

    Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

  4. Analysis of RA-8 critical facility core in some configurations; Analisis del nucleo de la facilidad critica RA-8 en distintas configuraciones

    Energy Technology Data Exchange (ETDEWEB)

    Abbate, Maximo J [Instituto de Investigaciones Cientificas y Tecnicas de las Fuerzas Armadas (CITEFA), Villa Martelli (Argentina); Sbaffoni, Maria M [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Gerencia de Tecnologia

    2000-07-01

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  5. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core

  6. Comfort-oriented vehicle suspension design with skyhook inerter configuration

    Science.gov (United States)

    Hu, Yinlong; Chen, Michael Z. Q.; Sun, Yonghui

    2017-09-01

    This paper is concerned with the comfort-oriented vehicle suspension design problem by using a skyhook inerter configuration. The rationale of the skyhook inerter is to use a grounded inerter to virtually increase the sprung mass of a vehicle, as it is analytically demonstrated that increasing the sprung mass can always improve the ride comfort performance. Semi-active means to realize the skyhook inerter configuration are investigated by using semi-active inerters. Three control laws, that is the on-off control, the anti-chatter on-off control, and the continuous control, are proposed for the semi-active inerter to approximate the skyhook inerter. Numerical simulations are performed to demonstrate the effectiveness and performances of these control laws. It is shown that the semi-active realizations of the skyhook inerter by using the proposed control laws can achieve over 10% improvement compared with the traditional strut, and similar performances are obtained for these control laws, with slight differences with respect to different static stiffnesses of the suspension system.

  7. Design comparisons of TRU burner cores with similar sodium void worth

    International Nuclear Information System (INIS)

    Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

    2001-01-01

    This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

  8. The materials challenge for LFR core design

    International Nuclear Information System (INIS)

    Grasso, Giacomo; Agostini, Pietro

    2013-01-01

    LFR share the main issues of all Fast Reactors, while presenting specific issues due to the use of lead as coolant. A number of constraints impairs the design of a LFR core, possibly resulting in a viability domain not exploitable for producing electricity in an efficient (hence economic) way. In particular, the most restrictive issues to be faced pend on the cladding. The selection of proper cladding materials provides the solution for the issues impairing the resistance of the cladding against stresses and irradiation effects. On the other hand, the protection of the cladding requires surface protections like oxide scales (passivation) or adherent layers (coating). Oxide scales seem not sufficient for a stable and effective protection of the base material. The application of adherent layers seems the only promising solution for protecting the cladding against corrosion. For the short term (i.e.: ALFRED), advanced 15/15Ti with coating is the reference solution for the cladding, allowing a core design complying with all the design constraints and goals. The candidate coatings are already being tested under irradiation to proceed towards qualification. In parallel, new base materials and/or coatings are presently under investigation. For the long term (i.e.: ELFR), the availability of such advanced materials/coatings might allow the extension of the viability domain towards higher and broader ranges (temperature, dpa, etc.), extending the fields of applications of LFRs and resulting in higher performances

  9. Multimedia foundations core concepts for digital design

    CERN Document Server

    Costello, Vic; Youngblood, Susan

    2012-01-01

    Understand the core concepts and skills of multimedia production and digital storytelling using text, graphics, photographs, sound, motion, and video. Then, put it all together using the skills that you have developed for effective project planning, collaboration, visual communication, and graphic design. Presented in full color with hundreds of vibrant illustrations, Multimedia Foundations trains you in the principles and skill sets common to all forms of digital media production, enabling you to create successful, engaging content, no matter what tools you are using. Companion website

  10. Multi-band phase shifter design using modified slotline configuration

    Science.gov (United States)

    Kulandhaisamy, Indhumathi; Rajendran, Dinesh Babu; Kanagasabai, Malathi; Gurusamy, Gunasekaran; Moorthy, Balaji; George, Jithila V.; Lawrance, Livya

    2017-01-01

    In this paper, an analog multiband phase shifter using slotline configuration is proposed. To implement the design, a pair of modified Split Ring Resonator (SRR) is employed. The periodic property of SRR provides multiband characteristics, whether the coupling slot gives the phase variations over the bands. The operation is well explained with an equivalent circuit model and its characteristics have been studied both in simulation and measurement. The prototype operates in 1.77-2.16, 3.5-3.97, 5.08-5.33, 6.43-6.93, and 8.01-8.59 GHz frequency bands which can be utilized for GSM, GPS, WLAN, C-band, and X-band applications, respectively.

  11. Design evaluation of emergency core cooling systems using Axiomatic Design

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Gyunyoung [Massachusetts Institute of Technology, Department of Mechanical Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)]. E-mail: gheo@mit.edu; Lee, Song Kyu [Korea Advanced Institute of Science and Technology, Department of Nuclear and Quantum Engineering, 373-1 Guseong-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2007-01-15

    In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies.

  12. Lessons Learned in Designing User-configurable Modular Robotics

    DEFF Research Database (Denmark)

    Lund, Henrik Hautop

    2013-01-01

    User-configurable robotics allows users to easily configure robotic systems to perform task-fulfilling behaviors as desired by the users. With a user configurable robotic system, the user can easily modify the physical and func-tional aspect in terms of hardware and software components of a robotic...... with the semi-autonomous com-ponents of the user-configurable robotic system in interaction with the given environment. Components constituting such a user-configurable robotic system can be characterized as modules in a modular robotic system. Several factors in the definition and implementation...

  13. Neutronic design of the XT-ADS core

    International Nuclear Information System (INIS)

    Van den Eynde, G.

    2007-01-01

    The EUROTRANS project is an integrated project in the 6th European Framework Program in the context of Partitioning and Transmutation. The objective of this project is the step-wise approach to a European Transmutation Demonstration. This project aims to deliver an advanced design of a small-scale Accelerator Driven System (ADS), XT-ADS, as well as the conceptual design of a European Facility for Industrial Transmutation (EFIT). The partners of this project accepted to use the MYRRHA Draft-2 design file as a starting basis for the design of the short-term XT-ADS demonstration machine. Instead of starting from a blank page, this allowed optimising an existing design towards the needs of XT-ADS, and this within the accepted limits of the safety requirements. Many options have been revisited and the framework is now set up. The main two objectives of the XT-ADS machine are the following: to demonstrate the feasibility of the ADS concept and to perform as a multi-purpose irradiation facility. Special attention is paid to the possibility of testing fuel dedicated to transmutation of minor actinides and long-life fission products. During the demonstration phase, the core will be loaded with MOX fuel in a clean core configuration. Since the XT-ADS must be a representative prototype, it has to operate at a reasonable power, a minimum of 50 MWth was set in the objectives. After this phase, the core will house In-Pile-Sections of different types for irradiating material samples, new types of fuel pins. We aim to be able to provide irradiation conditions that are close to EFIT conditions so XT-ADS can be used as a test-bed for EFIT parts

  14. Neutronic design of the RSG-GAS compact core without CIP

    International Nuclear Information System (INIS)

    Susilo, Jati; Kuntoro, Iman

    2002-01-01

    Improvement of the efficiency of reactor operation can be chivvied by some ways, such as, the uranium density of the fuel, loading pattern and configuration of core elements. The paper deals with determination of optimal configuration of the compact core with out CIP. Calculations were carried out by means of SRAC-PIJ module for cross section generation and SRAC-ASMBURN for core calculations. The optimal compact core obtained, showed that no-CIP compact core increase highest reactivity value about 0,84 % Δk/k and longest time operation about 1,19 time in the safety criteria that is power peaking factor less then 1,4 and margin control element worth less then volume in the first design that -2,2% Δk/k

  15. Neutronic design of the RSG-GAS compact core without CIP

    International Nuclear Information System (INIS)

    Jati-Susilo; Iman-Kuntoro

    2003-01-01

    Improvement of the efficiency of reactor operation can be achieved by some ways, such as, the uranium density of the fuel, loading pattern and configuration of core elements. The paper deals with determination of optimal configuration of the compact core with out CIP. Calculations were carried out by means of SRAC-PIJ module for cross section generation and SRAC-ASMBURN for core calculations. The optimal compact core obtained, showed that no-CIP compact core increase highest reactivity value about 1.06 % Δk/k and longest time operation about 1.19 time in the safety criteria that is power peaking factor less then 1.4 and margin control element worth less then value in the first design that -2.2% Δk/k

  16. Study in stationary state of the subcriticality of intermediate configurations of core in the reloading process of a BWR

    International Nuclear Information System (INIS)

    Hernandez, J.L.; Montes, J.L.; Perusquia, R.; Ortiz, J.J.

    2006-01-01

    In this work is carried out the simulation in three dimensions with the COREMASTER-PRESTO code, of the behavior of the reactor core in different stages of the change process of fuel assemblies. To carry out the simulation, this code requires of a database of nuclear parameters that includes those that can associate to the areas of an assemblies that they don't contain fuel and in its place there is moderator. These nuclear parameters are calculated with the AURORA-HELIOS-ZENITH-TABGEN system. One of the approaches that were carried out consisted on designing a 'water assemble', that is to say, an axial arrangement of 25 'water cells'. To obtain the appropriate 'water cell' its were carried out some selective test cases, since it presents in two cases the necessity to find an enough minimum value of fissile material for the correct execution of HELIOS, firstly, and later on COREMASTER-PRESTO. In the first case, the situation is solved when placing symmetrically 6 bars with natural uranium in the lateral areas of the cell; with that which the value of k inf of 0.1592 is obtained in the calculations with the HELIOS code in the cold condition to zero power (CZP), and 0% of vacuums. For the second case the cell includes symmetrically 28 bars with natural uranium, and the k inf value is 0.45290. These values are the maximum through the life of the 'cell.' As part of the activities that are developed during the fuel substitution, this the one of evaluating the subcriticality of the core each determined number of substitution movements. The obtained results when evaluating the k-effective in cold condition, in 5 different intermediate core configurations, as the loading process of the fuel advances are presented. To make the evaluation with CM-PRESTO in each configuration it was proceeded to complete the rest of the 444 assemblies with the one denominated 'water assemble'. In all the evaluated cases the subcriticality of the core was demonstrated in cold condition and with

  17. Core design and performance of small inherently safe LMRs

    International Nuclear Information System (INIS)

    Orechwa, Y.; Khalil, H.; Turski, R.B.; Fujita, E.K.

    1986-01-01

    Oxide and metal-fueled core designs at the 900 MWt level and constrained by a requirement for interchangeability are described. The physics parameters of the two cores studied here indicate that metal-fueled cores display attractive economic and safety features and are more flexible than are oxide cores in adapting to currently-changing deployment scenarios

  18. The Design and Analysis of Virtual Network Configuration for a Wireless Mobile ATM Network

    OpenAIRE

    Bush, Stephen F.

    1999-01-01

    This research concentrates on the design and analysis of an algorithm referred to as Virtual Network Configuration (VNC) which uses predicted future states of a system for faster network configuration and management. VNC is applied to the configuration of a wireless mobile ATM network. VNC is built on techniques from parallel discrete event simulation merged with constraints from real-time systems and applied to mobile ATM configuration and handoff. Configuration in a mobile network is a dyna...

  19. Configuration and engineering design of the ARIES-RS tokamak power plant

    International Nuclear Information System (INIS)

    Tillack, M.S.; Malang, S.; Waganer, L.; Wang, X.R.; Sze, D.K.; El-Guebaly, L.; Wong, C.P.C.; Crowell, J.A.; Mau, T.K.; Bromberg, L.

    1997-01-01

    ARIES-RS is a conceptual design study which has examined the potential of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. The design is a 1000 MWe, DT-burning fusion power plant based on the reversed-shear tokamak mode of plasma operation, and using moderately advanced engineering concepts such as lithium-cooled vanadium-alloy plasma-facing components. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. The ARIES-RS engineering design process emphasized the attainment of the top-level mission requirements developed in the early part of the study in a collaborative effort between the ARIES Team and representatives from U.S. electric utilities and industry. Major efforts were devoted to develop a credible configuration that allows rapid removal of full sectors followed by disassembly in the hot cells during plant operation. This was adopted as the only practical means to meet availability goals. Use of an electrically insulating coating for the self-cooled blanket and divertor provides a wide design window and simplified design. Optimization of the shield, which is one of the larger cost items, significantly reduced the power core cost by using ferritic steel where the power density and radiation levels are low. An additional saving is made by radial segmentation of the blanket, such that large segments can be reused. The overall tokamak configuration is described here, together with each of the major fusion power core components: the first-wall, blanket and shield; divertor; heating, current drive and fueling systems; and magnet systems. (orig.)

  20. In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

    International Nuclear Information System (INIS)

    Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

    2010-01-01

    According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

  1. Learn from the Core--Design from the Core

    Science.gov (United States)

    Ockerse, Thomas

    2012-01-01

    The current objective, object-oriented approach to design is questioned along with design education viewed as a job-oriented endeavor. Instead relational knowledge and experience in a holistic sense, both tacit and explicit, are valued along with an appreciation of the unique character of the student. A new paradigm for design education is…

  2. Major integration issues in evolving the configuration design space for the Aries-Cs compact stellarator power plant

    International Nuclear Information System (INIS)

    Raffray, A.R.; Malang, S.; El-Guebaly, L.; Ihli, T.; Najmabadi, F.; Wang, X.

    2006-01-01

    The Aries-Cs study has been launched with the goal of developing through physics and engineering optimization an attractive power plant concept based on a compact stellarator configuration. The first phase of the study involved scoping out different physics configurations including two and three field period options. The engineering effort during that phase aimed at scoping out maintenance schemes and power core designs best suited to a compact stellarator configuration. This led to a down selection of the most attractive blanket configurations and maintenance schemes for more detailed studies during the second phase of the study. This paper summarizes early results from the second phase of the Aries-Cs study with a particular emphasis on the engineering effort

  3. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-11-01

    The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

  4. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-09-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

  5. A reverse depletion method for pressurized water reactor core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.; Kin, Y.J.

    1986-01-01

    Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. The large numbers of burnable poison pins used to control the power peaking at the in-board fresh fuel positions have introduced an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of burnable poison pins in the fresh fuel. A new method for performing core design more suitable for low-leakage fuel management is reported. A procedure was developed that uses the wellknown ''Haling depletion'' to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and burnable poison distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provided for the maximum utility to PWR core reload design

  6. Computer-Aided Test Flow in Core-Based Design

    NARCIS (Netherlands)

    Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

    2000-01-01

    This paper copes with the test-pattern generation and fault coverage determination in the core based design. The basic core-test strategy that one has to apply in the core-based design is stated in this work. A Computer-Aided Test (CAT) flow is proposed resulting in accurate fault coverage of

  7. A core design study for 'zero-sodium-void-worth' cores

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Suzuki, Masao; Hill, R.N.

    1992-01-01

    Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

  8. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    International Nuclear Information System (INIS)

    Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

    2012-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

  9. Design basis reconstitution and configuration management of nuclear power plants

    International Nuclear Information System (INIS)

    Smith, P.R.; Cook, R.A.

    1990-01-01

    This paper presents a method by which all plants (those under construction, as well as those in operation) can benefit from a formal, step-by-step approach to data capture, storage, and retrieval for use throughout the plant life cycle. To demonstrate the viability and in-practice success of a single three-dimensional Configuration Management database, the specific capabilities of one such system are discussed in detail: the Plant Configuration Management System (PLANT/CMS). (orig.)

  10. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

  11. Design and Optimisation of Fuel Tanks for BWB Configurations

    Directory of Open Access Journals (Sweden)

    Goraj Zdobyslaw

    2016-12-01

    Full Text Available This paper describes assumptions, goals, methods, results and conclusions related to fuel tank arrangement of a flying wing passenger airplane configuration. A short overview of various fuel tank systems in use today of different types of aircraft is treated as a starting point for designing a fuel tank system to be used on very large passenger airplanes. These systems may be used to move fuel around the aircraft to keep the centre of gravity within acceptable limits, to maintain pitch and lateral balance and stability. With increasing aircraft speed, the centre of lift moves aft, and for trimming the elevator or trimmer must be used thereby increasing aircraft drag. To avoid this, the centre of gravity can be shifted by pumping fuel from forward to aft tanks. The lesson learnt from this is applied to minimise trim drag by moving the fuel along the airplane. Such a task can be done within coming days if we know the minimum drag versus CG position and weight value. The main part of the paper is devoted to wing bending moment distribution. A number of arrangements of fuel in airplane tanks are investigated and a scenario of refuelling - minimising the root bending moments - is presented. These results were obtained under the assumption that aircraft is in long range flight (14 hours, CL is constant and equal to 0.279, Specific Fuel Consumption is also constant and that overall fuel consumption is equal to 20 tons per 1 hour. It was found that the average stress level in wing structure is lower if refuelling starts from fuel tanks located closer to longitudinal plane of symmetry. It can influence the rate of fatigue.

  12. Review of advanced core designs for LMFBRs

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    1986-01-01

    It is a matter of great importance for the development of LMFBR to reduce its power cost to the level of the other power generating means. For this purpose, some ideas that use advanced core concepts to reduce LMFBR's power cost by improving its fuel cycle economics have recently been proposed. In this report, two hopeful ideas that use advanced core concepts: (1) Ultra Long Life Core (ULLC) - non-refueling over LMFBR power plant life; (2) Integral Fast Reactor (IFR) concept - metal fueled core and pyrometallurgical reprocessing; are picked up and their economical effect and technical probrems are investigated. (author)

  13. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Lily Suparlina; Tukiran Surbakti

    2014-01-01

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  14. Neutronic Design of KALIMER-600 Core with Moderator Rods

    International Nuclear Information System (INIS)

    Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

    2004-01-01

    Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

  15. Westinghouse Nuclear Core Design Training Center - a design simulator

    International Nuclear Information System (INIS)

    Altomare, S.; Pritchett, J.; Altman, D.

    1992-01-01

    The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

  16. Computer-Aided Test Flow in Core-Based Design

    NARCIS (Netherlands)

    Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

    2000-01-01

    This paper copes with the efficient test-pattern generation in a core-based design. A consistent Computer-Aided Test (CAT) flow is proposed based on the required core-test strategy. It generates a test-pattern set for the embedded cores with high fault coverage and low DfT area overhead. The CAT

  17. Optimal burnable poison utilization in PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

  18. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life VHTR Configurations: Designs, Advantages and Limitations

    International Nuclear Information System (INIS)

    Tsvetkov, Pavel V.

    2009-01-01

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  19. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life Bhr Configurations: Designs, Advantages and Limitations

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Pavel V. Tsvetkov

    2009-05-20

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  20. Design Report for the core design of the first core Mark-Ia of the SNR-300

    International Nuclear Information System (INIS)

    Stanculescu, A.

    1984-05-01

    The report describes the first core Mark-Ia of the SNR-300 reactor and its different assembly types with their operational strategy. Methods, criteria and results of the neutron physical, thermal hydraulic and core mechanical design of the whole core and its assemblies are presented

  1. Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    The nominal power of the reconstructed Dalat reactor is of 500 KW. After a meticulous preparation the Russian and Vietnamese teams have proceeded to the physical reactor start-up in November 1983 with the core configuration exempt from the neutron trap. The reactor has reached the physical criticality at 19h50 on 1 November 1983. The report delineates different steps of the start-up procedure. 2 refs., 3 figs., 7 tabs

  2. Basic criticality relations for gas core design

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1992-01-01

    Minimum critical fissile concentrations are calculated for U-233, U-235, Pu-239, and Am-242m mixed homogeneously with hydrogen at temperatures to 15,000K. Minimum critical masses of the same mixtures in a 1000 liter sphere are also calculated. It is shown that propellent efficiencies of a gas core fizzler engine using Am-242m as fuel would exceed those in a solid core engine as small as 1000L operating at 100 atmospheres pressure. The same would be true for Pu-239 and possibly U-233 at pressures of 1000 atm. or at larger volumes

  3. Core design and fuel rod analyses of a super fast reactor with high power density

    International Nuclear Information System (INIS)

    Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

  4. Criteria design of the CAREM 25 reactor's core: neutronic aspects

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    The criteria that guided the design, from the neutronic point of view, of the CAREM reactor's core were presented. The minimum set of objectives and general criteria which permitted the design of the particular systems constituting the CAREM 25 reactor's core is detailed and stated. (Author) [es

  5. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-01-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System is presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented. (author)

  6. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  7. GNPS 18-months fuel cycles core thermal hydraulic design

    International Nuclear Information System (INIS)

    Liu Changwen; Zhou Zhou

    2002-01-01

    GNPS begins to implement the 18-month fuel cycles from the initial annual reload at cycle 9, thus the initial core thermal hydraulic design is not valid any more. The new critical heat flux (CHF) correlation, FC, which is developed by Framatome, is used in the design, and the generalized statistical methodology (GSM) instead of the initial deterministic methodology is used to determine the DNBR design limit. As the AFA 2G and AFA 3G are mixed loaded in the transition cycle, it will result that the minimum DNBR in the mixed core is less than that of AFA 3G homogenous core, the envelop mixed core DNBR penalty is given. Consequently the core physical limit for mixed core and equilibrium cycles, and the new over temperature ΔT overpower ΔT are determined

  8. How Cultural Knowledge Shapes Core Design Thinking

    DEFF Research Database (Denmark)

    Clemmensen, Torkil; Ranjan, Apara; Bødker, Mads

    2018-01-01

    The growing trend of co-creation and co-design in cross-cultural design teams presents challenges for the design thinking process. We integrate two frameworks, one on reasoning patterns in design thinking, the other on the dynamic constructivist theory of culture, to propose a situation specific...... framework for the empirical analysis of design thinking in cross-cultural teams. We illustrate the framework with a qualitative analysis of 16 episodes of design related conversations, which are part of a design case study. The results show that cultural knowledge, either as shared by the cross......-cultural team or group specific knowledge of some team members, shape the reasoning patterns in the design thinking process across all the 16 episodes. Most of the design discussions were approached by the designers as problem situations that were formulated in a backward direction, where the value to create...

  9. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2009-01-01

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

  10. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

    2009-04-15

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

  11. In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

    International Nuclear Information System (INIS)

    Todt, W.H. Sr.; Playfoot, K.C.

    1988-01-01

    This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

  12. Core reset system design for linear induction accelerator

    International Nuclear Information System (INIS)

    Durga Praveen Kumar, D.; Mitra, S.; Sharma, Archana; Nagesh, K.V.; Chakravarthy, D.P.

    2006-01-01

    A repetitive pulsed power system based Linear Induction Accelerator (LIA-200) is being developed at BARC to get an electron beam of 200keV, 5kA, 50ns, 10-100 Hz. Amorphous core is the heart of these accelerators. It serves various functions in different subsystems viz. pulse power modulator, pulse transformer, magnetic switches and induction cavities. One of the factors that make the magnetic components compact is utilization of the total flux swing available in the core. In the present system, magnetic switches, pulse transformers, and induction cavity are designed to avail the full flux swing available in the core. For achieving this objective, flux density in the core has to be kept at the reverse saturation, before the main pulse is applied. The electrical circuit which makes it possible is called the core reset system. In this paper the details of core reset system designed for LIA-200 are described. (author)

  13. Using a Configuration System to Design Toilets and Place Installation Shafts

    DEFF Research Database (Denmark)

    Kudsk, Anders; Hvam, Lars; Thuesen, Christian Langhoff

    2013-01-01

    The aim of this research is to discover how configuration systems can support a product's design process when a high degree of variation is required and a very open or endless space exists for possible configurations. The article is based on an industrial case involving a firm that wishes to offer...... of opportunities exist for using configuration systems in the construction industry. They also show that it can be done without fundamentally changing the present process. © Kudsk et al.; Licensee Bentham Open....

  14. KALIMER database development (database configuration and design methodology)

    International Nuclear Information System (INIS)

    Jeong, Kwan Seong; Kwon, Young Min; Lee, Young Bum; Chang, Won Pyo; Hahn, Do Hee

    2001-10-01

    KALIMER Database is an advanced database to utilize the integration management for Liquid Metal Reactor Design Technology Development using Web Applicatins. KALIMER Design database consists of Results Database, Inter-Office Communication (IOC), and 3D CAD database, Team Cooperation system, and Reserved Documents, Results Database is a research results database during phase II for Liquid Metal Reactor Design Technology Develpment of mid-term and long-term nuclear R and D. IOC is a linkage control system inter sub project to share and integrate the research results for KALIMER. 3D CAD Database is s schematic design overview for KALIMER. Team Cooperation System is to inform team member of research cooperation and meetings. Finally, KALIMER Reserved Documents is developed to manage collected data and several documents since project accomplishment. This report describes the features of Hardware and Software and the Database Design Methodology for KALIMER

  15. Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

    Energy Technology Data Exchange (ETDEWEB)

    Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

    2017-02-15

    The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

  16. Coupled-core fluxgate magnetometer: Novel configuration scheme and the effects of a noise-contaminated external signal

    International Nuclear Information System (INIS)

    Palacios, Antonio; Aven, John; In, Visarath; Longhini, Patrick; Kho, Andy; Neff, Joseph D.; Bulsara, Adi

    2007-01-01

    Recent theoretical and experimental work has shown that unidirectional coupling can induce oscillations in overdamped and undriven nonlinear dynamical systems that are non-oscillatory when uncoupled; in turn, this has been shown to lead to new mechanisms for weak (compared to the energy barrier height) signal detection and amplification. The potential applications include fluxgate magnetometers, electric field sensors, and arrays of Superconducting Quantum Interference Device (SQUID) rings. In the particular case of the fluxgate magnetometer, we have developed a ''coupled-core fluxgate magnetometer'' (CCFM); this device has been realized in the laboratory and its dynamics used to quantify many properties that are generic to this class of systems and coupling. The CCFM operation is underpinned by the emergent oscillatory behavior in a unidirectionally coupled ring of wound ferromagnetic cores, each of which can be treated as an overdamped bistable dynamic system when uncoupled. In particular, one can determine the regimes of existence and stability of the (coupling-induced) oscillations, and the scaling behavior of the oscillation frequency. More recently, we studied the effects of a (Gaussian) magnetic noise floor on a CCFM system realized with N=3 coupled ferromagnetic cores. In this Letter, we first introduce a variation on the basic CCFM configuration that affords a path to enhanced device sensitivity, particularly for N>=3 coupled elements. We then analyze the response of the basic CCFM configuration as well as the new setup to a dc target signal that has a small noisy component (or ''contamination'')

  17. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

    2008-07-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  18. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2008-01-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  19. Establishment of Systematic Design Control/Configuration Management Processes to Enhance Engineering Capability

    International Nuclear Information System (INIS)

    Inagaki, T.; Hamada, T.; Ihara, T.

    2016-01-01

    Full text: After the accident of Fukushima Daiichi Nuclear Power Plant in 2011, Tokyo Electric Power Company (TEPCO) launched various measures to enhance plant safety and safety culture of its employees. One of the important aspects of these measures is to enhance engineering capability and TEPCO is conducting actions to establish systematic design control and configuration management processes as an important foundation of such engineering capability. This paper describes how TEPCO is establishing systematic configuration management processes from three aspects, i.e., design requirement and bases management, facility configuration control, and configuration change management. It also provides brief information of the IT systems that are being introduced and will support the systematic design control and configuration management processes. (author

  20. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

    2002-03-01

    In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

  1. Fundamental attributes of a practical configuration management program for nuclear plant design control

    International Nuclear Information System (INIS)

    Klein, S.M.

    1988-06-01

    This summarizes the results of an evaluation of findings identifies during a number of Safety-System Functional Inspections and Safety System Outage Modification Inspections which are related to configuration management for nuclear plant design control. A computerized database of these findings was generated from a review of the design inspection reports. Based on the results of the evaluation, attributes of a configuration management program were developed which are responsive to minimizing these types of inspection findings. Incorporation of these key attributes is considered good practice in the development of a configuration management program for design control at operating nuclear plants

  2. [Three-dimensional computer aided design for individualized post-and-core restoration].

    Science.gov (United States)

    Gu, Xiao-yu; Wang, Ya-ping; Wang, Yong; Lü, Pei-jun

    2009-10-01

    To develop a method of three-dimensional computer aided design (CAD) of post-and-core restoration. Two plaster casts with extracted natural teeth were used in this study. The extracted teeth were prepared and scanned using tomography method to obtain three-dimensional digitalized models. According to the basic rules of post-and-core design, posts, cores and cavity surfaces of the teeth were designed using the tools for processing point clouds, curves and surfaces on the forward engineering software of Tanglong prosthodontic system. Then three-dimensional figures of the final restorations were corrected according to the configurations of anterior teeth, premolars and molars respectively. Computer aided design of 14 post-and-core restorations were finished, and good fitness between the restoration and the three-dimensional digital models were obtained. Appropriate retention forms and enough spaces for the full crown restorations can be obtained through this method. The CAD of three-dimensional figures of the post-and-core restorations can fulfill clinical requirements. Therefore they can be used in computer-aided manufacture (CAM) of post-and-core restorations.

  3. Adjoint-Based Aerodynamic Design of Complex Aerospace Configurations

    Science.gov (United States)

    Nielsen, Eric J.

    2016-01-01

    An overview of twenty years of adjoint-based aerodynamic design research at NASA Langley Research Center is presented. Adjoint-based algorithms provide a powerful tool for efficient sensitivity analysis of complex large-scale computational fluid dynamics (CFD) simulations. Unlike alternative approaches for which computational expense generally scales with the number of design parameters, adjoint techniques yield sensitivity derivatives of a simulation output with respect to all input parameters at the cost of a single additional simulation. With modern large-scale CFD applications often requiring millions of compute hours for a single analysis, the efficiency afforded by adjoint methods is critical in realizing a computationally tractable design optimization capability for such applications.

  4. Development of underwater high-definition camera for the confirmation test of core configuration and visual examination of BWR fuel

    International Nuclear Information System (INIS)

    Watanabe, Masato; Tuji, Kenji; Ito, Keisuke

    2010-01-01

    The purpose of this study is to develop underwater High-Definition camera for the confirmation test of core configuration and visual examination of BWR fuels in order to reduce the time of these tests and total cost regarding to purchase and maintenance. The prototype model of the camera was developed and examined in real use condition in spent fuel pool at HAMAOKA-2 and 4. The examination showed that the ability of prototype model was either equaling or surpassing to conventional product expect for resistance to radiation. The camera supposes to be used in the dose rate condition of under about 10 Gy/h. (author)

  5. Reactor Core Design and Analysis for a Micronuclear Power Source

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

    Full Text Available Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

  6. Scalable Multi-core Architectures Design Methodologies and Tools

    CERN Document Server

    Jantsch, Axel

    2012-01-01

    As Moore’s law continues to unfold, two important trends have recently emerged. First, the growth of chip capacity is translated into a corresponding increase of number of cores. Second, the parallalization of the computation and 3D integration technologies lead to distributed memory architectures. This book provides a current snapshot of industrial and academic research, conducted as part of the European FP7 MOSART project, addressing urgent challenges in many-core architectures and application mapping.  It addresses the architectural design of many core chips, memory and data management, power management, design and programming methodologies. It also describes how new techniques have been applied in various industrial case studies. Describes trends towards distributed memory architectures and distributed power management; Integrates Network on Chip with distributed, shared memory architectures; Demonstrates novel design methodologies and frameworks for multi-core design space exploration; Shows how midll...

  7. Core design with respect to the safety concept

    International Nuclear Information System (INIS)

    Kollmar, W.

    1981-01-01

    In the present paper the following topics are dealt with: Principles of reactor core design and optimization, fuel management and safety concept for higher cycles and results of risk analyses (e.g. rod ejection, steam line break etc.) (RW)

  8. CopperCore, an Open Source IMS Learning Design Engine

    NARCIS (Netherlands)

    Vogten, Hubert

    2004-01-01

    The presentation gives an overview of the approach of the development programme of the OTEC department towards the development of Open Source. The CopperCore IMS Learning Design engine is described as an example of this approach.

  9. Gamow-Teller response in the configuration space of a density-functional-theory-rooted no-core configuration-interaction model

    Science.gov (United States)

    Konieczka, M.; Kortelainen, M.; Satuła, W.

    2018-03-01

    Background: The atomic nucleus is a unique laboratory in which to study fundamental aspects of the electroweak interaction. This includes a question concerning in medium renormalization of the axial-vector current, which still lacks satisfactory explanation. Study of spin-isospin or Gamow-Teller (GT) response may provide valuable information on both the quenching of the axial-vector coupling constant as well as on nuclear structure and nuclear astrophysics. Purpose: We have performed a seminal calculation of the GT response by using the no-core configuration-interaction approach rooted in multireference density functional theory (DFT-NCCI). The model treats properly isospin and rotational symmetries and can be applied to calculate both the nuclear spectra and transition rates in atomic nuclei, irrespectively of their mass and particle-number parity. Methods: The DFT-NCCI calculation proceeds as follows: First, one builds a configuration space by computing relevant, for a given physical problem, (multi)particle-(multi)hole Slater determinants. Next, one applies the isospin and angular-momentum projections and performs the isospin and K mixing in order to construct a model space composed of linearly dependent states of good angular momentum. Eventually, one mixes the projected states by solving the Hill-Wheeler-Griffin equation. Results: The method is applied to compute the GT strength distribution in selected N ≈Z nuclei including the p -shell 8Li and 8Be nuclei and the s d -shell well-deformed nucleus 24Mg. In order to demonstrate a flexibility of the approach we present also a calculation of the superallowed GT β decay in doubly-magic spherical 100Sn and the low-spin spectrum in 100In. Conclusions: It is demonstrated that the DFT-NCCI model is capable of capturing the GT response satisfactorily well by using a relatively small configuration space, exhausting simultaneously the GT sum rule. The model, due to its flexibility and broad range of applicability, may

  10. Improvement of SSR core design for ABWR-II

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Okada, Hiroyuki; Kitamura, Hideya; Sakurada, Koichi; Tanabe, Akira

    2003-01-01

    In order to enhance the spectral shift effect in the ABWR-II reactor, a novel core design to bring out better performance of spectral shift rods (SSRs) is studied. The SSR is a new type of water rod, in which the water level develops naturally during operation and changes according to the coolant flow rate through the channel. By using the SSR, the average moderator density, which is directly related to core reactivity, can be controlled over a wide range by the core flow rate. In the new SSR core design, two types of SSR bundles, in which settings for the SSR water levels are different, are utilized and loaded according to flow distribution in the core. This two-region SSR core design allows wide variation in the average SSR water level, thus improving fuel economy. Enhancement of SSR function in the two-region SSR core increases the uranium saving factor by about 25%, from the 6% of the conventional uniform SSR core to about 8%. (author)

  11. Preliminary conceptual design of target system. Pt. 1. System configuration

    Energy Technology Data Exchange (ETDEWEB)

    Hino, Ryutaro; Haga, Katsuhiro; Kaminaga, Masanori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1997-07-01

    In the 21st century, neutron is expected to play a very important role in the fields of structural biology, nuclear physics, material science if a very high-intensity neutron source will be built because of its superior nature as an probe to investigate material structure and its function. The Japan Atomic Energy Research Institute has launched the Neutron Science Project for construction and utilization of a high-intensity spallation neutron source coupled with a proton accelerator. In the project, a neutron scattering facility is planned to be constructed in an early stage. Development of a 5MW spallation neutron source is one of the most difficult technical challenges in this project. A two-step development plan of the target was established to construct a 5MW-target station In the 1st step, a 1.5MW target will be constructed to develop 5MW target technology. The preliminary conceptual design was conducted to clarify the specifications of the target system of 1.5MW and 5MW including system layout, scale etc. This report describes (1) a design policy, (2) a layout of system consisting of the target, remote-handling devices, bio-shieldings etc., (3) specifications of components and facilities such as cooling systems for target and moderators, beam-port shutter and air conditioning system, (4) overhaul procedures by remote-handling devices, (5) safety assessment, and (6) necessary R and D items derived from the design activity. (author)

  12. Study on core design for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Okubo, Tsutomu

    2002-01-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  13. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  14. Review on JMTR safety design for LEU core conversion

    International Nuclear Information System (INIS)

    Komori, Yoshihiro; Yokokawa, Makoto; Saruta, Toru; Inada, Seiji; Sakurai, Fumio; Yamamoto, Katsumune; Oyamada, Rokuro; Saito, Minoru

    1993-12-01

    Safety of the JMTR was fully reviewed for the core conversion to low enriched uranium fuel. Fundamental policies for the JMTR safety design were reconsidered based on the examination guide for safety design of test and research reactors, and safety of the JMTR was confirmed. This report describes the safety design of the JMTR from the viewpoint of major functions for reactor safety. (author)

  15. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  16. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

  17. Configuration Method Design for Reconfigurable Manufacturing System with the aid of Plant Simulation

    DEFF Research Database (Denmark)

    Li, Yang; Zhang, Shuai; Bilberg, Arne

    2014-01-01

    A new Reconfigurable Manufacturing System structure has been recently designed by a large consumer goods manufacturer in Europe, aiming to balance the performance of productivity and flexibility. This article shows an exploratory research on the (re)configuration procedure of the new RMS structure....... Following the procedure which is designed in this paper, the (re)configuration of RMS can be managed as part of the daily operation with the help of computer simulation. Keywords: Plant Simulation, Tecnomatix, Reconfigurable Manufacturing System, modular manufacturing....

  18. Practical Consideration Factors to Design Array Configuration of Direction Finding System for Airborne Signal Intelligence

    Directory of Open Access Journals (Sweden)

    Jong-Hwan Lee

    2018-01-01

    Full Text Available Airborne signal intelligence (SIGINT systems must be capable of locating radio signal sources. Direction finding (DF to support this capability is an important factor. There are some practical considerations to be taken when designing the array configuration of a DF system for airborne SIGINT systems. This paper summarizes the practical factors when designing the array configuration of the DF system for airborne SIGINT. In particular, it focuses on four areas: antenna consideration factors when installing the DF system for airborne SIGINT from a practical point of view, array configuration methods for airborne communications intelligence and electronic intelligence, and a numerical analysis to select the optimum antenna position for airborne SIGINT.

  19. Mars Rover Sample Return aerocapture configuration design and packaging constraints

    Science.gov (United States)

    Lawson, Shelby J.

    1989-01-01

    This paper discusses the aerodynamics requirements, volume and mass constraints that lead to a biconic aeroshell vehicle design that protects the Mars Rover Sample Return (MRSR) mission elements from launch to Mars landing. The aerodynamic requirements for Mars aerocapture and entry and packaging constraints for the MRSR elements result in a symmetric biconic aeroshell that develops a L/D of 1.0 at 27.0 deg angle of attack. A significant problem in the study is obtaining a cg that provides adequate aerodynamic stability and performance within the mission imposed constraints. Packaging methods that relieve the cg problems include forward placement of aeroshell propellant tanks and incorporating aeroshell structure as lander structure. The MRSR missions developed during the pre-phase A study are discussed with dimensional and mass data included. Further study is needed for some missions to minimize MRSR element volume so that launch mass constraints can be met.

  20. Forming Core Elements for Strategic Design Management

    DEFF Research Database (Denmark)

    Beim, Anne; Vibæk, Kasper Sánchez

    2007-01-01

    . This evolution also leads to a demand for precise definitions of the values and qualities that can be used as managing tools in common building practice and it puts the traditional architectural design process under pressure. This paper outlines an approach to architectural quality as dealt with in the design...... and how the architectural potentials are realized when dealing with modern industrial processes are examined. To analyse and structure the empirical data, a model was developed consisting of four approaches for action. The approaches are categorized along different dichotomies in order to point out...... the model was tested in the architectural education at the Royal Danish Academy of Fine Arts - School of Architecture. The overall research project has two aims - to help offices identify the characteristics and specific methods of working with architectural quality in an industrialized context...

  1. A Minimum Shuffle Core Design Strategy for ESBWR

    International Nuclear Information System (INIS)

    Karve, A.A.; Fawcett, R.M.

    2008-01-01

    The Economic Simplified Boiling Water Reactor (ESBWR) is GEH's next evolution of advanced BWR technology. There are 1132 fuel bundles in the core and the thermal power is 4500 MWt. Similar to conventional plants there is an outage after a specified period of operation, when the plant shuts down. During the outage a specified fraction of fuel bundles are discharged from the core, it is loaded with the same fraction of fresh fuel, and fuel is shuffled to obtain an optimum core design that meets the goals for a successful operation of the next cycle. The discharge, load, and the associated shuffles are time-consuming and expensive tasks that impact the overall outage schedule and costs. Therefore, there is an incentive to keep maneuvers to a minimum and to perform them more efficiently. The benefits for a large core, such as the ESBWR with 1132 fuel bundles, are escalated. This study focuses on a core reload design strategy to minimize the total number of shuffles during an outage. A traditional equilibrium cycle is used as a reference basis, which sets the reference number of shuffles. In the minimum shuffle core design however, a set of two equilibrium cycles (N and N+1, referred to as a 'bi- equilibrium' cycle) is envisioned where the fresh fuel of cycle N (that becomes the once-burnt fuel of cycle N+1) ideally does not move in the two cycles. The cost of fuel efficiency is determined for obtaining such a core loading by comparing it to the traditional equilibrium cycle. There are several additional degrees of freedom when designing a bi-equilibrium cycle that could be utilized, and the potential benefits of these flexibilities are assessed. In summary, the feasibility of a minimum shuffle fuel cycle and core design for an ESBWR is studied. The cost of fuel efficiency is assessed in comparison to the traditional design. (authors)

  2. Automated Design and Optimization of Pebble-bed Reactor Cores

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Ougouag, Abderrafi M.; Terry, William K.

    2010-01-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  3. Simulation of Molten Core-Concrete Interaction in oxide/metal stratified configuration with the TOLBIAC-ICB code

    International Nuclear Information System (INIS)

    Tourniaire, B.; Spindler, B.

    2005-01-01

    The frame of this work is the validation of the TOLBIAC-ICB code which is devoted to the simulation of Molten Core-Concrete Interaction (MCCI) for reactor safety analysis. Attention focuses here on the validation of TOLBIAC-ICB in configurations expected to be representative of the long term phase of MCCI i.e. during an interaction between an oxide/metal stratified corium melt and a concrete structure. Up to now the BETA tests performed at the Forschungszentrum Karlsruhe (FzK) are the only tests available to study such kind of interaction. The BETA tests are first described and the operating conditions are reminded. The TOLBIAC-ICB code is then briefly described, with emphasis on the models used for stratified configurations. The results of the simulations are discussed. A sensitivity study is also performed with the power generated in the oxide layer instead of the metal layer as in the test. This last calculation shows that the large axial ablation observed in the tests is probably due to the peculiar configuration of the test with input power in the bottom metal layer. Since in the reactor case the residual power would be mainly concentrated in the upper oxide layer, the conclusions of the BETA tests for the reactor applications, in term of axial ablation, must be derived with caution. (author)

  4. A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR (Sodium-cooled Fast Reactor) cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.

  5. Coupled-core fluxgate magnetometer: Novel configuration scheme and the effects of a noise-contaminated external signal

    Energy Technology Data Exchange (ETDEWEB)

    Palacios, Antonio [San Diego State University, Nonlinear Dynamical Systems Group, Department of Mathematics and Statistics, San Diego, CA 92182-7720 (United States)]. E-mail: palacios@euler.sdsu.edu; Aven, John [San Diego State University, Nonlinear Dynamical Systems Group, Department of Mathematics and Statistics, San Diego, CA 92182-7720 (United States); In, Visarath [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States)]. E-mail: visarath@spawar.navy.mil; Longhini, Patrick [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States); Kho, Andy [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States); Neff, Joseph D. [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States); Bulsara, Adi [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States)]. E-mail: bulsara@spawar.navy.mil

    2007-07-16

    Recent theoretical and experimental work has shown that unidirectional coupling can induce oscillations in overdamped and undriven nonlinear dynamical systems that are non-oscillatory when uncoupled; in turn, this has been shown to lead to new mechanisms for weak (compared to the energy barrier height) signal detection and amplification. The potential applications include fluxgate magnetometers, electric field sensors, and arrays of Superconducting Quantum Interference Device (SQUID) rings. In the particular case of the fluxgate magnetometer, we have developed a ''coupled-core fluxgate magnetometer'' (CCFM); this device has been realized in the laboratory and its dynamics used to quantify many properties that are generic to this class of systems and coupling. The CCFM operation is underpinned by the emergent oscillatory behavior in a unidirectionally coupled ring of wound ferromagnetic cores, each of which can be treated as an overdamped bistable dynamic system when uncoupled. In particular, one can determine the regimes of existence and stability of the (coupling-induced) oscillations, and the scaling behavior of the oscillation frequency. More recently, we studied the effects of a (Gaussian) magnetic noise floor on a CCFM system realized with N=3 coupled ferromagnetic cores. In this Letter, we first introduce a variation on the basic CCFM configuration that affords a path to enhanced device sensitivity, particularly for N>=3 coupled elements. We then analyze the response of the basic CCFM configuration as well as the new setup to a dc target signal that has a small noisy component (or ''contamination'')

  6. Analysis of Wolsong-1 SDS1 Effectiveness with Stuck-In Shutoff Rod Core Configurations

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Jin; Jung, Young Suk; Choi, Seong Soo [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sung Min [Korea Hydro and Nuclear Power Co., Ltd., Seoul (Korea, Republic of)

    2010-05-15

    The Wolsong-1 CANDU 6 reactor (W-1) is currently undergoing the major refurbishment project including replacement of the pressure tube after nearly 25 years of service. In parallel to the refurbishment, the reactor is planned to be operated with Improved Technical Specifications (ITS) that are being prepared as an integrated part of the new project to conduct the overall Improved Standard Technical Specifications (ISTS) layout for PHWR (Ref. 1). The ISTS project is dually purported, namely, firstly, to improve and update the existing Current Technical Specifications (CTS) with the specific emphasis of rooting the conceptual and practical applications that are derived out of the PWR oriented TS so that PHWR could be operated in more closely surveillant practices with PWR domestically, and secondly, the finished ISTS product could also be exposed overseas for global marketing purposes. During the course of reviewing the draft version of the W-1 ITS it is felt that ITS Items related to the unavailability of Shutdown System No. 1 (SDS1) should be supported with some detailed analysis performed by using the safety analysis codes as a precautionary measure. The present paper deals with the cases of SDS1 shutoff rod (SOR) stuck into the core so that the stuck rod will not be available when SDS1 is actuated to drop rods into the core. In the following, the models used for the simulations are briefly described and the corresponding results are presented with some conclusions.

  7. Design and development of small and medium integral reactor core

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR's, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs

  8. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  9. GFR fuel and core pre-conceptual design studies

    International Nuclear Information System (INIS)

    Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M.; Malo, J.Y.; Garnier, J.C.; Bertrand, F.; Bosq, J.C.

    2007-01-01

    The revision of the GFR core design - plate type - has been undertaken since previous core presented at Global'05. The self-breeding searched for has been achieved with an optimized design ('12/06 E'). The higher core pressure drop was a matter of concern. First of all, the core coolability in natural circulation for pressurized conditions has been studied and preliminary plant transient calculations have been performed. The design and safety criteria are met but no more margin remains. The project is also addressing the feasibility and the design of the fuel S/A. The hexagonal shape together with the principle of closed S/A (wrapper tube) is kept. Ceramic plate type fuel element combines a high enough core power density (minimization of the Pu inventory) and plutonium and minor actinides recycling capabilities. Innovative for many aspects, the fuel element is central to the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is considered as the other fuel element of interest. This combination of fuel/core feasibility and performance analysis, safety dispositions and performances analysis will compose the 'GFR preliminary feasibility' which is a project milestone at the end of the year 2007. (authors)

  10. VLSI Architecture for Configurable and Low-Complexity Design of Hard-Decision Viterbi Decoding Algorithm

    Directory of Open Access Journals (Sweden)

    Rachmad Vidya Wicaksana Putra

    2016-06-01

    Full Text Available Convolutional encoding and data decoding are fundamental processes in convolutional error correction. One of the most popular error correction methods in decoding is the Viterbi algorithm. It is extensively implemented in many digital communication applications. Its VLSI design challenges are about area, speed, power, complexity and configurability. In this research, we specifically propose a VLSI architecture for a configurable and low-complexity design of a hard-decision Viterbi decoding algorithm. The configurable and low-complexity design is achieved by designing a generic VLSI architecture, optimizing each processing element (PE at the logical operation level and designing a conditional adapter. The proposed design can be configured for any predefined number of trace-backs, only by changing the trace-back parameter value. Its computational process only needs N + 2 clock cycles latency, with N is the number of trace-backs. Its configurability function has been proven for N = 8, N = 16, N = 32 and N = 64. Furthermore, the proposed design was synthesized and evaluated in Xilinx and Altera FPGA target boards for area consumption and speed performance.

  11. Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

    International Nuclear Information System (INIS)

    Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

    2001-01-01

    The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

  12. Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

    2010-01-01

    The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

  13. Preliminary Core Design Analysis of a 200MWth Pebble Bed-type VHTR

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Noh, Jae Man

    2007-01-01

    This paper intends to suggest the preliminary core design analysis of a VHTR for a hydrogen production. The nuclear hydrogen system that utilizes the high temperature heat generated from the VHTR is a promising candidate for a cost effective, safe and clean supply of hydrogen in the age of hydrogen economy. Among two candidate VHTR cores, that is, a prismatic modular reactor (PMR) and a pebble bed-type reactor (PBR), we focus on the design of a 200MWth PBR (hereinafter PBR200) in this paper. Here, the 200MWth power is selected for a demonstration plant. The core configuration of the PBR200 is similar to the PBMR (Pebble Bed Modular Reactor, 400MWth) of South Africa, but the overall dimension of the reactor system is scaled-down. This paper is to suggest two candidate PBR200 cores. One is an annular core with an inner reflector (PBR200-CD1) which was presented at IWRES07, and the other is a cylindrical core without an inner reflector (PBR200-CD2)

  14. Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Zhou, E-mail: zhaozhou@swip.ac.cn; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

    2016-02-15

    Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li{sub 4}SiO{sub 4} pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

  15. Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

    International Nuclear Information System (INIS)

    Zhao, Zhou; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

    2016-01-01

    Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li_4SiO_4 pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

  16. Data on the configuration design of internet-connected home cooling systems by engineering students.

    Science.gov (United States)

    McComb, Christopher; Cagan, Jonathan; Kotovsky, Kenneth

    2017-10-01

    This experiment was carried out to record the step-by-step actions that humans take in solving a configuration design problem, either in small teams or individually. Specifically, study participants were tasked with configuring an internet-connected system of products to maintain temperature within a home, subject to cost constraints. Every participant was given access to a computer-based design interface that allowed them to construct and assess solutions. The interface was also used to record the data that is presented here. In total, data was collected for 68 participants, and each participant was allowed to perform 50 design actions in solving the configuration design problem. Major results based on the data presented here have been reported separately, including initial behavioral analysis (McComb et al.) [1], [2] and design pattern assessments via Markovian modeling (McComb et al., 2017; McComb et al., 2017) [3], [4].

  17. A Core Design Approach Aimed at Sustainability and Intrinsic Safety

    International Nuclear Information System (INIS)

    Grasso, Giacomo

    2013-01-01

    The comprehensive approach adopted for the core design of all LFRs investigated within the LEADER project, proved to effectively drive the design to the fulfillment of the aimed sustainability performances, and the respect of the design constraints for the robust implementation of the inherent safety principle: • the ELFR core is able to operate adiabatically, with a very narrow reactivity swing along a 2.5 y cycle; • wide margins are provided for protecting the fuel and the structures even in case of unprotected transients, allowing for very long grace times

  18. Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

  19. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  20. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  1. Refractory metal component technology for in-core sensor design

    International Nuclear Information System (INIS)

    Cannon, C.P.

    1986-02-01

    Within recent years, an increasing concern over reactor safety has prompted tests that characterize reactor core environments during transient conditions. Such tests include the Loss-of-Fluid-Tests (Idaho National Engineering Lab (INEL)), Severe Fuel Damage Tests (INEL), Core Debris Rubble Tests (Sandia National Laboratories (SNL)), and similar tests performed by foreign nations. The in-core sensors for these tests require refractory metal components to be compatible with electrical insulator materials as well as materials comprising highly corrosive service mediums. This paper presents the refractory metal technology utilized to provide basic sensor designs in the above mentioned reactor tests

  2. Ex-vessel core catcher design requirements and preliminary concepts evaluation

    International Nuclear Information System (INIS)

    Friedland, A.J.; Tilbrook, R.W.

    1974-01-01

    As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

  3. Characteristics of fast reactor core designs and closed fuel cycle

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

    2007-01-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

  4. Core compressor exit stage study. Volume 3: Data and performance report for screening test configurations

    Science.gov (United States)

    Wisler, D. C.

    1980-01-01

    Rear stage blading designs that have lower losses in their endwall boundary layer regions were developed. Test data and performance results for rotor B, stator B, and stator C - blading designs that offer promise of reducing endwall losses relative to the baseline are given. A low speed research compressor was the principal investigative tool. The tests were conducted using four identical stages of blading so that the test data would be obtained in a true multistage environment.

  5. Design of Computerized in Core Fuel Management System of Kartini Reactor

    International Nuclear Information System (INIS)

    Edi-Trijono-Budisantoso; Sardjono, Y; Edi-Purwanto; Widi-Setiawan

    2000-01-01

    The program organization for managing Kartini reactor fuel elements has been designed. This program organization work to process on-line operationdata-base and core configuration data-base to produce data-base for in-corefuer management. The in-core fuel management data-base consist of irradiationhistory card, radionuclides inventory and radiation dose for each fuelelement. The computation in this process based on the ORIGEN2, TRIGAP codesand some in-house developed codes that perform matching between output dataof many codes to output data of other code. This program organization worksunder control of a program manager by following the scheduled time table. Thedesign gives a description of the first step development of the in-core fuelmanagement that will be implemented in the internet web server. (author)

  6. Core Design Concept and Core Structural Material Development for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint (~0.4MPa) are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test

  7. A Configurable Surface-Electrode Ion Trap Design for Quantum Information Processing

    International Nuclear Information System (INIS)

    Liu Wei; Chen Shu-Ming; Chen Ping-Xing; Wu Wei

    2013-01-01

    We propose a configurable surface-electrode ion trap design to alleviate the poor reusability of the existing traps. It can architecturally and electrically support 5 mainstream modes by design reuse, thus enhancing the trap reusability and reducing the experiment setup overhead. We also develop a corresponding simulation suite which can optimize trap geometries and calculate trap parameters to control the trapped ion's classic motion. According to our analytical and simulated results, the configurable design can serve as a unified platform for basic research of large-scale quantum information processing

  8. Core design of super LWR with double tube water rods

    International Nuclear Information System (INIS)

    Wu, Jianhui; Oka, Yoshiaki

    2014-01-01

    Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

  9. Soft shell hard core concept for aircraft impact resistant design

    International Nuclear Information System (INIS)

    Chen, C.; Rieck, P.J.

    1978-01-01

    For nuclear power plants sited in the vicinity of airports, the hypothetical events of aircraft impact have to be designed for. The conventional design concept is to strengthen the exterior structure to resist the impact induced force. The stiffened structures have two (2) disadvantages; one is the high construction cost, and the other is the high reaction force induced as well as the vibrational effects on the interior equipment and piping systems. This new soft shell hard core concept can relieve the above shortcomings. In this concept, the essential equipment required for safety are installed inside the hard core area for protection and the non-essential equipment are maintained between the hard core and soft shell area. During a hypothetical impact event, the soft shell will collapse locally and absorb large amounts of kinetic energy; hence, it reduces the reaction force and the vibrational effects. The design and analysis of the soft shell concept are discussed. (Author)

  10. Status of core nuclear design technology for future fuel

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

    1997-01-01

    The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

  11. The APR1400 Core Design by Using APA Code System

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Koh, Byung Marn

    2008-01-01

    The nuclear design for APR1400 has been performed to prepare the core model for Automatic Load Follow Operation Simulation. APA (ALPHA/ PHOENIXP/ ANC) code system is a tool for the multi-cycle depletion calculations for APR1400. Its detail versions for ALPHA, PHOENIX-P and ANC are 8.9.3, 8.6.1 and 8.10.5, respectively. The first and equilibrium core depletion calculations for APR1400 have been performed to assure the target cycle length and confirm the safety parameters. The parameters are satisfied within limitation about nuclear design criteria. This APR1400 core models will be based on the design parameters for APR1400 Simulator

  12. Design and economic implications of heterogeneity in an LMFBR core

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1983-01-01

    Much emphasis is currently being placed in LMFBR design on reducing both the capital cost and the fuel cycle cost of an LMFBR to insure its economic competativeness without a rapid increase in the uranium prices. In this study the relationship between two core design options, their neutronic consequences, and their effect on fuel cycle cost are analyzed. The two design options are the selection of pin diameter and the degree of heterogeneity. In the case of a heterogeneous core, with a low sodium void reactivity worth this ratio of fertile internal blanket to driver assemblies is generally about 0.40. However, some advantages of cores with heterogeneity of 0.08 to 0.2 for a fixed pin diameter have been reported

  13. Core design concepts for high performance light water reactors

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2007-01-01

    Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

  14. Benefits of Low Boron Core Design Concept for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2009-10-15

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in {sup 10}B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts.

  15. Benefits of Low Boron Core Design Concept for PWR

    International Nuclear Information System (INIS)

    Daing, Aung Tharn; Kim, Myung Hyun

    2009-01-01

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in 10 B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts

  16. VG-400 atomic power and technological installation. Possible core design

    International Nuclear Information System (INIS)

    Komarov, E.V.; Laptev, F.V.; Lyubivyj, A.G.; Mitenkov, F.M.; Samojlov, O.B.; Sukhachevskij, Yu.B.

    1979-01-01

    The main characteristics, basic circuit and configuration of equipment of the VG-400 atomic power and technological installation are considered. This installation is intended for supplying with highly-potential heat of thermal electrochemical hydrogen production and for power generation in the steam-turbine cycle. The main installation characteristics: HTGR reactor heat power 1100 MW, electric power 300 MW, helium coolant pressure 50 atm, output temperature 950 deg C, steam pressure in the second contour 175 atm, temperature 535 deg C, core diameter and height 6.4 m and 4 m, respectively, number of spherical fuel elements 8.5x10 5 . The installation can ensure hydrogen production of 10 5 Nxm 3 /h. For the VG-400 reactor block the integral arrangement of the first circuit equipment in the reinforced concrete is chosen. Two versions of the reactor core with prismatic and spherical fuel elements are compared. It is shown that taking into account great potentialities of the spherical zone in a case of further temperature increase and its positive qualities with respect to construction and processing of fuel elements and graphite blocks, the utilization of simplier units and mechanisms in the overloading system and in the process of profiling of energy distribution the choice of the spherical configuration for the VG-400 pilot plant installation seems to be valid

  17. Core design methodology and software for Temelin NPP

    International Nuclear Information System (INIS)

    Havluj, F; Hejzlar, J.; Klouzal, J.; Stary, V.; Vocka, R.

    2011-01-01

    In the frame of the process of fuel vendor change at Temelin NPP in the Czech Republic, where, starting since 2010, TVEL TVSA-T fuel is loaded instead of Westinghouse VVANTAGE-6 fuel, new methodologies for core design and core reload safety evaluation have been developed. These documents are based on the methodologies delivered by TVEL within the fuel contract, and they were further adapted according to Temelin NPP operational needs and according to the current practice at NPP. Along with the methodology development the 3D core analysis code ANDREA, licensed for core reload safety evaluation in 2010, have been upgraded in order to optimize the safety evaluation process. New sequences of calculations were implemented in order to simplify the evaluation of different limiting parameters and output visualization tools were developed to make the verification process user friendly. Interfaces to the fuel performance code TRANSURANUS and sub-channel analysis code SUBCAL were developed as well. (authors)

  18. Status of experimental data for the VHTR core design

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-05-01

    The VHTR (Very High Temperature Reactor) is being emerged as a next generation nuclear reactor to demonstrate emission-free nuclear-assisted electricity and hydrogen production. The VHTR could be either a prismatic or pebble type helium cooled, graphite moderated reactor. The final decision will be made after the completion of the pre-conceptual design for each type. For the pre-conceptual design for both types, computational tools are being developed. Experimental data are required to validate the tools to be developed. Many experiments on the HTGR (High Temperature Gas-cooled Reactor) cores have been performed to confirm the design data and to validate the design tools. The applicability and availability of the existing experimental data have been investigated for the VHTR core design in this report.

  19. Wechsler Adult Intelligence Scale-Revised Block Design broken configuration errors in nonpenetrating traumatic brain injury.

    Science.gov (United States)

    Wilde, M C; Boake, C; Sherer, M

    2000-01-01

    Final broken configuration errors on the Wechsler Adult Intelligence Scale-Revised (WAIS-R; Wechsler, 1981) Block Design subtest were examined in 50 moderate and severe nonpenetrating traumatically brain injured adults. Patients were divided into left (n = 15) and right hemisphere (n = 19) groups based on a history of unilateral craniotomy for treatment of an intracranial lesion and were compared to a group with diffuse or negative brain CT scan findings and no history of neurosurgery (n = 16). The percentage of final broken configuration errors was related to injury severity, Benton Visual Form Discrimination Test (VFD; Benton, Hamsher, Varney, & Spreen, 1983) total score and the number of VFD rotation and peripheral errors. The percentage of final broken configuration errors was higher in the patients with right craniotomies than in the left or no craniotomy groups, which did not differ. Broken configuration errors did not occur more frequently on designs without an embedded grid pattern. Right craniotomy patients did not show a greater percentage of broken configuration errors on nongrid designs as compared to grid designs.

  20. PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

  1. Design assumptions and bases for small D-T-fueled Sperical Tokamak (ST) fusion core

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Galambos, J.D.; Fogarty, P.J.

    1996-01-01

    Recent progress in defining the assumptions and clarifying the bases for a small D-T-fueled ST fusion core are presented. The paper covers several issues in the physics of ST plasmas, the technology of neutral beam injection, the engineering design configuration, and the center leg material under intense neutron irradiation. This progress was driven by the exciting data from pioneering ST experiments, a heightened interest in proof-of-principle experiments at the MA level in plasma current, and the initiation of the first conceptual design study of the small ST fusion core. The needs recently identified for a restructured fusion energy sciences program have provided a timely impetus for examining the subject of this paper. Our results, though preliminary in nature, strengthen the case for the potential realism and attractiveness of the ST approach

  2. Core designs for the de-regulated market

    International Nuclear Information System (INIS)

    Almberger, J.; Bernro, R.; Pettersson, H.

    1999-01-01

    Complete text of publication follows: The electricity market deregulation in the Nordic countries encourages innovations and cost reductions for power production in the Vattenfall reactors. The competition on the electricity market is strong, electricity price reductions dramatic and uncertainties about the future power demand is large. In the fuel area this situation has given increased attention to traditional areas like flexibility in power production, improved core designs, need for margins (improved fuel designs), improved surveillance, decreased lead times. At Vattenfall new fuel designs are already being implemented following the last fuel purchase, for which flexibility and margins, were given high values in the evaluations with the multipurpose task of eliminating fuel related problems and meeting the future market situation. This strategy has given Vattenfall a flying start to meeting the demands of the de-regulated market. What has been added are broad studies undertaken to investigate the various route into the future with respect to finding the most effective strategies for fuel and core design and optimization. In the present paper the Vattenfall priorities for fuel designs and margins are presented in a schematic manner summarizing the results of the last fuel purchase and also presenting the current program for LFAs. Technical limitations, licensing and R and D aspects, with respect to improving the fuel utilization will be mentioned. The main focus in the paper is on the broad study carried out in the PWR core design area. Driven by the relatively low power demand various possibilities for higher production flexibility have been investigated specifically extended coast-down, coast-up and yearly load follow. Further to reduce the costs for fuel consumption improvements in core designs have been studied: improved low leakage loading patterns, low enriched end zones, improved Gd designs etc. Main results and conclusions of the core design studies will

  3. High Performance Systolic Array Core Architecture Design for DNA Sequencer

    Directory of Open Access Journals (Sweden)

    Saiful Nurdin Dayana

    2018-01-01

    Full Text Available This paper presents a high performance systolic array (SA core architecture design for Deoxyribonucleic Acid (DNA sequencer. The core implements the affine gap penalty score Smith-Waterman (SW algorithm. This time-consuming local alignment algorithm guarantees optimal alignment between DNA sequences, but it requires quadratic computation time when performed on standard desktop computers. The use of linear SA decreases the time complexity from quadratic to linear. In addition, with the exponential growth of DNA databases, the SA architecture is used to overcome the timing issue. In this work, the SW algorithm has been captured using Verilog Hardware Description Language (HDL and simulated using Xilinx ISIM simulator. The proposed design has been implemented in Xilinx Virtex -6 Field Programmable Gate Array (FPGA and improved in the core area by 90% reduction.

  4. Applying CBR to machine tool product configuration design oriented to customer requirements

    Science.gov (United States)

    Wang, Pengjia; Gong, Yadong; Xie, Hualong; Liu, Yongxian; Nee, Andrew Yehching

    2017-01-01

    Product customization is a trend in the current market-oriented manufacturing environment. However, deduction from customer requirements to design results and evaluation of design alternatives are still heavily reliant on the designer's experience and knowledge. To solve the problem of fuzziness and uncertainty of customer requirements in product configuration, an analysis method based on the grey rough model is presented. The customer requirements can be converted into technical characteristics effectively. In addition, an optimization decision model for product planning is established to help the enterprises select the key technical characteristics under the constraints of cost and time to serve the customer to maximal satisfaction. A new case retrieval approach that combines the self-organizing map and fuzzy similarity priority ratio method is proposed in case-based design. The self-organizing map can reduce the retrieval range and increase the retrieval efficiency, and the fuzzy similarity priority ratio method can evaluate the similarity of cases comprehensively. To ensure that the final case has the best overall performance, an evaluation method of similar cases based on grey correlation analysis is proposed to evaluate similar cases to select the most suitable case. Furthermore, a computer-aided system is developed using MATLAB GUI to assist the product configuration design. The actual example and result on an ETC series machine tool product show that the proposed method is effective, rapid and accurate in the process of product configuration. The proposed methodology provides a detailed instruction for the product configuration design oriented to customer requirements.

  5. Applying design principles to fusion reactor configurations for propulsion in space

    International Nuclear Information System (INIS)

    Carpenter, S.A.; Deveny, M.E.; Schulze, N.R.

    1993-01-01

    The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. Three design principles (DP's) were applied to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. A preliminary rating of these configurations was performed, and it was concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS)

  6. Neutronic design of mixed oxide-silicide cores for the core conversion of rsg-gas reactor

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Tukiran; Pinem surian; Febrianto

    2001-01-01

    The core conversion of rsg-gas reactor from an all-oxide (U 3 O 8 -Al) core, through a series of mixed oxide-silicide core, to an all-silicide (U 3 Si 2 -Al) core for the same meat density of 2.96 g U/cc is in progress. The conversion is first step of the step-wise conversion and will be followed by the second step that is the core conversion from low meat density of silicide core, through a series of mixed lower-higher density of silicide core, to an all-higher meat density of 3.55 g/cc core. Therefore, the objectives of this work is to design the mixed cores on the neutronic performance to achieve safety a first full-silicide core for the reactor with the low uranium meat density of 2.96gU/cc. The neutronic design of the mixed cores was performed by means of Batan-EQUIL-2D and Batan-3DIFF computer codes for 2 and 3 dimension diffusion calculation, respectively. The result shows that all mixed oxide-silicide cores will be feasible to achieve safety a fist full-silicide core. The core performs the same neutronic core parameters as those of the equilibrium silicide core. Therefore, the reactor availability and utilization during the core conversion is not changed

  7. Economic evaluation of flying-qualities design criteria for a transport configured with relaxed static stability

    Science.gov (United States)

    Sliwa, S. M.

    1980-01-01

    Direct constrained parameter optimization was used to optimally size a medium range transport for minimum direct operating cost. Several stability and control constraints were varied to study the sensitivity of the configuration to specifying the unaugmented flying qualities of transports designed to take maximum advantage of relaxed static stability augmentation systems. Additionally, a number of handling qualities related design constants were studied with respect to their impact on the design.

  8. New design on air-core resistive NMR imaging magnet

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yan; Mingwu, Fan; Yixin, Miao

    1984-08-01

    A new type of NMR imaging air-core resistive magnet is designed. Based on the BIM Magnetostatic calculation the resultant four equiradial coils structure with optimized shapes of cross section possesses a larger spherical working volume obviously, comparing with the common four-coils imaging magnet. The manufacturing tolerance is also calculated.

  9. 300 MWe Burner Core Design with two Enrichment Zoning

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Sang Ji; Kim, Yeong Il

    2008-01-01

    KAERI has been developing the KALIMER-600 core design with a breakeven fissile conversion ratio. The core is loaded with a ternary metallic fuel (TRU-U-10Zr), and the breakeven characteristics are achieved without any blanket assembly. As an alternative plan, a KALIMER-600 burner core design has been also performed. In the early stage of the development of a fast reactor, the main purpose is an economical use of a uranium resource but nowadays in addition to the maximum utilization of a uranium resource, the burning of a high level radioactive waste is taken as an additional interest for the harmony of the environment. In way of constructing the commercial size reactor which has the power level ranging from 800 MWe to 1600 MWe, the demonstration reactor which has the power level ranging from 200 MWe to 600 MWe was usually constructed for the midterm stage to commercial size reactor. In this paper, a 300 MWe burner core design was performed with purpose of demonstration reactor for KALIMER-600 burner of 600 MWe. As a means to flatten the power distribution, instead of a single fuel enrichment scheme adapted in design of KALIMER-600 burner, the 2 enrichment zoning approach was adapted

  10. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  11. Westinghouse loading pattern search methodology for complex core designs

    International Nuclear Information System (INIS)

    Chao, Y.A.; Alsop, B.H.; Johansen, B.J.; Morita, T.

    1991-01-01

    Pressurized water reactor core designs have become more complex and must meet a plethora of design constraints. Trends have been toward longer cycles with increased discharge burnup, increased burnable absorber (BA) number, mixed BA types, reduced radial leakage, axially blanketed fuel, and multiple-batch feed fuel regions. Obtaining economical reload core loading patterns (LPs) that meet design criteria is a difficult task to do manually. Automated LP search tools are needed. An LP search tool cannot possibly perform an exhaustive search because of the sheer size of the combinatorial problem. On the other hand, evolving complexity of the design features and constraints often invalidates expert rules based on past design experiences. Westinghouse has developed a sophisticated loading pattern search methodology. This methodology is embodied in the LPOP code, which Westinghouse nuclear designers use extensively. The LPOP code generates a variety of LPs meeting design constraints and performs a two-cycle economic evaluation of the generated LPs. The designer selects the most appropriate patterns for fine tuning and evaluation by the design codes. This paper describes the major features of the LPOP methodology that are relevant to fulfilling the aforementioned requirements. Data and examples are also provided to demonstrate the performance of LPOP in meeting the complex design needs

  12. FIREBIRD - a conceptual design of a field reversed configuration compact torus fusion reactor (CTFR)

    International Nuclear Information System (INIS)

    Raman, R.; Zubrin, R.M.

    1987-01-01

    This paper is a summary of the work carried out by the Nuclear Engineering 512 design team at the University of Washington on a conceptual design study of a Compact-Torus (Field-Reversed) Fusion Reactor Configuration (CTFR). The primary objective of the study was to develop a reactor design that strived for high engineering power density, modest recirculating power and competitive cost of electrical power. A Conceptual design was developed for a translating field-reversed configuration reactor; based on the Physics developed by Tuszewski and Lindford at LANL and by Hoffman and Milroy at MSNW. Furthermore, it also appears possible to operate a simplified form of this reactor using a pure D-D fuel cycle after an initial D-T ignition ramp to reach the advanced fuel operating regime. One optimistic reactor so designed has a length of about 35 meters, producing a net electrical power of about 375 MWe

  13. ASTRID core: Design objectives, design approach, and R&D in support

    International Nuclear Information System (INIS)

    Mignot, G.; Devictor, N.

    2012-01-01

    ASTRID core design is mainly guided by safety objectives: 1. Prevention of the core meltdown accident: To prevent meltdown accidents: - by a natural behavior of the core and the reactor (no actuation of the two shutdown systems); - with adding passive complementary systems if natural behavior is not sufficient for some transient cases. 2. Mitigation of the fusion accident: To garantee that core fusion accidents don’t lead to significant mechanical energy release, whatever initiator event: - by a natural core behavior; - with adding specific mitigation dispositions in case of natural behavior is not suffficient

  14. The Design and Analysis of Virtual Network Configuration for a Wireless Mobile ATM Network

    Science.gov (United States)

    Bush, Stephen F.

    1999-05-01

    This research concentrates on the design and analysis of an algorithm referred to as Virtual Network Configuration (VNC) which uses predicted future states of a system for faster network configuration and management. VNC is applied to the configuration of a wireless mobile ATM network. VNC is built on techniques from parallel discrete event simulation merged with constraints from real-time systems and applied to mobile ATM configuration and handoff. Configuration in a mobile network is a dynamic and continuous process. Factors such as load, distance, capacity and topology are all constantly changing in a mobile environment. The VNC algorithm anticipates configuration changes and speeds the reconfiguration process by pre-computing and caching results. VNC propagates local prediction results throughout the VNC enhanced system. The Global Positioning System is an enabling technology for the use of VNC in mobile networks because it provides location information and accurate time for each node. This research has resulted in well defined structures for the encapsulation of physical processes within Logical Processes and a generic library for enhancing a system with VNC. Enhancing an existing system with VNC is straight forward assuming the existing physical processes do not have side effects. The benefit of prediction is gained at the cost of additional traffic and processing. This research includes an analysis of VNC and suggestions for optimization of the VNC algorithm and its parameters.

  15. GCRA review and appraisal of HTGR reactor-core-design program

    International Nuclear Information System (INIS)

    1980-09-01

    The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

  16. Ocean Thermal Energy Conversion (OTEC) platform configuration and integration. Volume II. Conceptual design. Final report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-01

    The purpose of this project is to evaluate six candidate hullforms as candidates for the OTEC commercial plant. This volume is a summary of the conceptual design including facility requirements, cost, schedule, and site sensitivity. Two OTEC commercial plant configurations are considered in this study: the ship and the semi-submersible. Engineering drawings are presented. (WHR)

  17. Design of a smart textile mat to study pressure distribution on multiple foam material configurations

    NARCIS (Netherlands)

    Donselaar, van R.; Chen, W.

    2011-01-01

    In this paper, we present a design of a smart textile pressure mat to study the pressure distribution with multiple foam material configurations for neonatal monitoring at Neonatal Intensive Care Units (NICU). A smart textile mat with 64 pressure sensors has been developed including software at the

  18. Organizational design configurations in the early stages of firm’s life cycle

    OpenAIRE

    Mosca, Luigi; Gianecchini, Martina; Campagnolo, Diego

    2017-01-01

    http://ocs.editorial.upv.es/index.php/CARMA/CARMA2016 Mosca, L.; Gianecchini, M.; Campagnolo, D. (2017). Organizational design configurations in the early stages of firm’s life cycle. Editorial Universitat Politècnica de València. doi:10.4995/CARMA2016.2015.3104. OCS

  19. Intelligent system for conceptural design of new reactor cores

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki

    1995-01-01

    The software system IRDS has been developed at Japan Atomic Energy Research Institute to support the conceptual design of a new type of reactor core in the fields of neutronics, thermohydraulics, and fuel behavior. IRDS involves various analysis codes, database, and man-machine interfaces that efficiently support a whole design process on a computer. The main purpose of conceptual design is to decide an optimal set of basic design parameters. Designers usually carry out many parametric survey calculations and search a design window (DW), which is a feasible parameter range satisfying design criteria and goals. An automatic DW search function is installed to support such works. The man-machine interface based on menu windows will enable nonspecialists to use various analysis codes easily

  20. Adaption of the PARCS Code for Core Design Audit Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyong Chol; Lee, Young Jin; Uhm, Jae Beop; Kim, Hyunjik [Nuclear Safety Evaluation, Daejeon (Korea, Republic of); Jeong, Hun Young; Ahn, Seunghoon; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The eigenvalue calculation also includes quasi-static core depletion analyses. PARCS has implemented variety of features and has been qualified as a regulatory audit code in conjunction with other NRC thermal-hydraulic codes such as TRACE or RELAP5. In this study, as an adaptation effort for audit applications, PARCS is applied for an audit analysis of a reload core design. The lattice physics code HELIOS is used for cross section generation. PARCS-HELIOS code system has been established as a core analysis tool. Calculation results have been compared on a wide spectrum of calculations such as power distribution, critical soluble boron concentration, and rod worth. A reasonable agreement between the audit calculation and the reference results has been found.

  1. Core design study on reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Hiroshi, Akie; Yoshihiro, Nakano; Toshihisa, Shirakawa; Tsutomu, Okubo; Takamichi, Iwamura

    2002-01-01

    The conceptual core design study of reduced-moderation water reactors (RMWRs) with tight-pitched MOX-fuelled lattice has been carried out at JAERI. Several different RMWR core concepts based on both BWR and PWR have been proposed. All the core concepts meet with the aim to achieve both a conversion ratio of 1.0 or larger and negative void reactivity coefficient. As one of these RMWR concepts, the ABWR compatible core is also proposed. Although the conversion ratio of this core is 1.0 and the void coefficient is negative, the discharge burn-up of the fuel was about 25 GWd/t. By adopting a triangular fuel pin lattice for the reduction of moderator volume fraction and modifying axial Pu enrichment distribution, it was aimed to extend the discharge burn-up of ABWR compatible type RMWR. By using a triangular fuel lattice of smaller moderator volume fraction, discharge burn-up of 40 GWd/t seems achievable, keeping the high conversion ratio and the negative void coefficient. (authors)

  2. Design of a flight director/configuration management system for piloted STOL approaches

    Science.gov (United States)

    Hoh, R. H.; Klein, R. H.; Johnson, W. A.

    1973-01-01

    The design and characteristics of a flight director for V/STOL aircraft are discussed. A configuration management system for piloted STOL approaches is described. The individual components of the overall system designed to reduce pilot workload to an acceptable level during curved, decelerating, and descending STOL approaches are defined. The application of the system to augmentor wing aircraft is analyzed. System performance checks and piloted evaluations were conducted on a flight simulator and the results are summarized.

  3. Improving Battery Reactor Core Design Using Optimization Method

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2011-01-01

    The Battery Omnibus Reactor Integral System (BORIS) is a small modular fast reactor being designed at Seoul National University to satisfy various energy demands, to maintain inherent safety by liquid-metal coolant lead for natural circulation heat transport, and to improve power conversion efficiency with the Modular Optimal Balance Integral System (MOBIS) using the supercritical carbon dioxide as working fluid. This study is focused on developing the Neutronics Optimized Reactor Analysis (NORA) method that can quickly generate conceptual design of a battery reactor core by means of first principle calculations, which is part of the optimization process for reactor assembly design of BORIS

  4. Design and analysis of PCRV core cavity closure

    International Nuclear Information System (INIS)

    Lee, T.T.; Schwartz, A.A.; Koopman, D.C.A.

    1980-05-01

    Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction

  5. Overview of PEC core design and requirements for PEC core restraint systems

    International Nuclear Information System (INIS)

    Cecchini, F.

    1984-01-01

    The Italian PEC reactor is an experimental loop type fast reactor of 120 MW thermal. Its main purpose is the in-pile development of fast reactor fuel. The mechanical principles in PEC core design and current modifications to ensure a safe seismic perturbation and shutdown are discussed in this paper. These anti-seismic modifications are aimed to limit the extent of reactivity perturbation during the seismic event and to guarantee control rod entry at any time during the seismic event

  6. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  7. Phononic band gap design in honeycomb lattice with combinations of auxetic and conventional core

    International Nuclear Information System (INIS)

    Mukherjee, Sushovan; Gopalakrishnan, S; Fabrizio Scarpa

    2016-01-01

    We present a novel design of a honeycomb lattice geometry that uses a seamless combination of conventional and auxetic cores, i.e. elements showing positive and negative Poisson’s ratio. The design is aimed at tuning and improving the band structure of periodic cellular structures. The proposed cellular configurations show a significantly wide band gap at much lower frequencies compared to their pure counterparts, while still retaining their major dynamic features. Different topologies involving both auxetic inclusions in a conventional lattice and conversely hexagonal cellular inclusions in auxetic butterfly lattices are presented. For all these cases the impact of the varying degree of auxeticity on the band structure is evaluated. The proposed cellular designs may offer significant advantages in tuning high-frequency bandgap behaviour, which is relevant to phononics applications. The configurations shown in this paper may be made iso-volumetric and iso-weight to a given regular hexagonal topology, making possible to adapt the hybrid lattices to existing sandwich structures with fixed dimensions and weights. This work also features a comparative study of the wave speeds corresponding to different configurations vis-a vis those of a regular honeycomb to highlight the superior behaviour of the combined hybrid lattice. (paper)

  8. Designing the colorectal cancer core dataset in Iran

    Directory of Open Access Journals (Sweden)

    Sara Dorri

    2017-01-01

    Full Text Available Background: There is no need to explain the importance of collection, recording and analyzing the information of disease in any health organization. In this regard, systematic design of standard data sets can be helpful to record uniform and consistent information. It can create interoperability between health care systems. The main purpose of this study was design the core dataset to record colorectal cancer information in Iran. Methods: For the design of the colorectal cancer core data set, a combination of literature review and expert consensus were used. In the first phase, the draft of the data set was designed based on colorectal cancer literature review and comparative studies. Then, in the second phase, this data set was evaluated by experts from different discipline such as medical informatics, oncology and surgery. Their comments and opinion were taken. In the third phase refined data set, was evaluated again by experts and eventually data set was proposed. Results: In first phase, based on the literature review, a draft set of 85 data elements was designed. In the second phase this data set was evaluated by experts and supplementary information was offered by professionals in subgroups especially in treatment part. In this phase the number of elements totally were arrived to 93 numbers. In the third phase, evaluation was conducted by experts and finally this dataset was designed in five main parts including: demographic information, diagnostic information, treatment information, clinical status assessment information, and clinical trial information. Conclusion: In this study the comprehensive core data set of colorectal cancer was designed. This dataset in the field of collecting colorectal cancer information can be useful through facilitating exchange of health information. Designing such data set for similar disease can help providers to collect standard data from patients and can accelerate retrieval from storage systems.

  9. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  10. Neutronic design of the RSG-GAS silicide core

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Kuntoro, I.; Hastowo, H. [Center for Development of Research Reactor Technology National Nuclear Energy Agency BATAN, PUSPIPTEK Serpong Tangerang, 15310 (Indonesia)

    2002-07-01

    The objective of core conversion program of the RSG-GAS multipurpose reactor is to convert the fuel from oxide, U{sub 3}O{sub 8}-Al to silicide, U{sub 3}Si{sub 2}-Al. The aim of the program is to gain longer operation cycle by having, which is technically possible for silicide fuel, a higher density. Upon constraints of the existing reactor system and utilization, an optimal fuel density in amount of 3.55 g U/cc was found. This paper describes the neutronic parameter design of the silicide equilibrium core and the design of its transition cores as well. From reactivity control point of view, a modification of control rod system is also discussed. All calculations are carried out by means of diffusion codes, Batan-EQUIL-2D, Batan-2DIFF and -3DIFF. The silicide core shows that longer operation cycle of 32 full power days can be achieved without decreasing the safety criteria and utilization capabilities. (author)

  11. Core compressor exit stage study. 1: Aerodynamic and mechanical design

    Science.gov (United States)

    Burdsall, E. A.; Canal, E., Jr.; Lyons, K. A.

    1979-01-01

    The effect of aspect ratio on the performance of core compressor exit stages was demonstrated using two three stage, highly loaded, core compressors. Aspect ratio was identified as having a strong influence on compressors endwall loss. Both compressors simulated the last three stages of an advanced eight stage core compressor and were designed with the same 0.915 hub/tip ratio, 4.30 kg/sec (9.47 1bm/sec) inlet corrected flow, and 167 m/sec (547 ft/sec) corrected mean wheel speed. The first compressor had an aspect ratio of 0.81 and an overall pressure ratio of 1.357 at a design adiabatic efficiency of 88.3% with an average diffusion factor or 0.529. The aspect ratio of the second compressor was 1.22 with an overall pressure ratio of 1.324 at a design adiabatic efficiency of 88.7% with an average diffusion factor of 0.491.

  12. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap; Khoi dong vat ly lo phan ung hat nhan Da Lat voi cau hinh vung hoat co bay notron

    Energy Technology Data Exchange (ETDEWEB)

    Hien, Pham Duy; Huy, Ngo Quang; Long, Vu Hai; Mai, Tran Khanh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs.

  13. Hyper-heuristic applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    Domingos, R P; Platt, G M

    2013-01-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  14. Simulated annealing algorithm for reactor in-core design optimizations

    International Nuclear Information System (INIS)

    Zhong Wenfa; Zhou Quan; Zhong Zhaopeng

    2001-01-01

    A nuclear reactor must be optimized for in core fuel management to make full use of the fuel, to reduce the operation cost and to flatten the power distribution reasonably. The author presents a simulated annealing algorithm. The optimized objective function and the punishment function were provided for optimizing the reactor physics design. The punishment function was used to practice the simulated annealing algorithm. The practical design of the NHR-200 was calculated. The results show that the K eff can be increased by 2.5% and the power distribution can be flattened

  15. MOX - equilibrium core design and trial irradiation in KAPS - 1

    International Nuclear Information System (INIS)

    Pradhan, A.S.; Ray, Sherly; Kumar, A.N.; Parikh, M.V.

    2006-01-01

    Option of usage of MOX fuel bundles in the equilibrium core of Indian 220 MWe PHWRs on a regular basis has been studied. The design of the MOX bundle considered is MOX -7 with inner 7 elements with uranium and plutonium oxide MOX fuel and outer 12 elements with natural uranium fuel. The composition of the plutonium isotopes corresponds to that at about 6500 MWD/TeU burnup. Burnup optimization has been done such that operation at design rated power is possible while achieving the maximum average discharge burnup. Operation with the optimized burnup pattern will result in substantial saving of natural uranium bundles. To obtain feedback on the performance of MOX bundles prior to its large scale use about 50 MOX-7 bundles have been loaded in KAPS - 1 equilibrium core. Locations have been selected such that reactor should be operating at rated power without violating any constraints on channel bundle powers and also meeting the safety requirements. Burnup of interest also should be achieved in minimum period of time. The fissile plutonium content in the 50 MOX fuel bundles loaded is about 75.6 wt % . About 38 bundles out of the 50 bundles loaded have been already discharged and remaining bundles are still in the core. The maximum discharge burnup of the MOX bundles is about 12000 MWD/TeU. The performance of the MOX bundles were excellent and as per prediction. No MOX bundle is reported to be failed. (author)

  16. Preliminary design of the new Proton Synchrotron Internal Dump core

    CERN Document Server

    AUTHOR|(CDS)2091975; Nuiry, François-Xavier

    The luminosity of the LHC particle accelerator at CERN is planned to be upgraded in the first half of 2020s, requiring also the upgrade of its injector accelerators, including the Proton Synchrotron (PS). The PS Internal Dumps are beam dumps located in the PS accelerator ring. They are safety devices designed to stop the circulating proton beam in order to protect the accelerator from damage due to an uncontrolled beam loss. The PS Internal Dumps need to be upgraded to be able to withstand the future higher intensity and energy proton beams. The dump core is a block of material interacting with the beam. It is located in ultra-high vacuum and moved into the beam path in 150 milliseconds by an electromagnet and spring-based actuation mechanism. The circulating proton beam is shaved by the core surface during thousands of beam revolutions. The preliminary new dump core design weighs 13 kilograms and consists of an isostatically pressed fine-grain graphite and a precipitation hardened copper alloy CuCrZr. The ...

  17. ATM Technology Demonstration-1 Phase II Boeing Configurable Graphical Display (CGD) Software Design Description

    Science.gov (United States)

    Wilber, George F.

    2017-01-01

    This Software Description Document (SDD) captures the design for developing the Flight Interval Management (FIM) system Configurable Graphics Display (CGD) software. Specifically this SDD describes aspects of the Boeing CGD software and the surrounding context and interfaces. It does not describe the Honeywell components of the CGD system. The SDD provides the system overview, architectural design, and detailed design with all the necessary information to implement the Boeing components of the CGD software and integrate them into the CGD subsystem within the larger FIM system. Overall system and CGD system-level requirements are derived from the CGD SRS (in turn derived from the Boeing System Requirements Design Document (SRDD)). Display and look-and-feel requirements are derived from Human Machine Interface (HMI) design documents and working group recommendations. This Boeing CGD SDD is required to support the upcoming Critical Design Review (CDR).

  18. Beams configuration design in target area with successive quadratic programming method

    International Nuclear Information System (INIS)

    Shi Zhiquan; Tan Jichun; Wei Xiaofeng; Man Jongzai; Zhang Xiaomin; Yuan Jing; Yuan Xiaodong

    1998-01-01

    The author describes the application of successive quadratic programming method (SQP) to design laser beam configuration in target area. Based on the requirement of ICF experiment physics, a math model of indirect-driver beam geometry is given. A 3D wire-frame is plotted, in which support lines represent 60 laser entireties and 240 turning points of support lines' segments stand for the spatial positions of reflectors

  19. Developing engineering design core competences through analysis of industrial products

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

    Most product development work carried out in industrial practice is characterised by being incremental, i.e. the industrial company has had a product in production and on the market for some time, and now time has come to design a new and upgraded variant. This type of redesign project requires...... that the engineering designers have core design competences to carry through an analysis of the existing product encompassing both a user-oriented side and a technical side, as well as to synthesise solution proposals for the new and upgraded product. The authors of this paper see an educational challenge in staging...... a course module, in which students develop knowledge, understanding and skills, which will prepare them for being able to participate in and contribute to redesign projects in industrial practice. In the course module Product Analysis and Redesign that has run for 8 years we have developed and refined...

  20. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  1. LFR core design for prevention & mitigation of severe accidents

    International Nuclear Information System (INIS)

    Grasso, Giacomo

    2012-01-01

    Conclusions: • Aiming at fully complying Gen-IV safety requirements – even in case of Fukushima-like events –, prevention and mitigation strategies must be stressed in FR design. • The safety of Lead-cooled Fast Reactors can rely on intrinsic features due to the coolant, such as: • the practical impossibility of Lead boiling, hence the unreliability of core (only) voiding for wide safety margins, and the retention of corium; • the high density of lead, for the buoyancy of Control Rods (allowing their safe positioning below the core), and the dispersion of molten core up to the setting up of a “cold melting pot”. • the possibility to adopt wide coolant channels for encouraging natural circulation, without affecting the hardness of the neutron spectrum; • the hard neutron spectrum allows the adiabatic operation of LFRs (which implies minimal criticality swings even through long cycles) with small amounts of Mas (hence with a negligible detriment to the safety features); • an effective reduction of the coolant density effect simply through the shortening of the active height

  2. Space Launch System, Core Stage, Structural Test Design and Implementation

    Science.gov (United States)

    Shaughnessy, Ray

    2017-01-01

    As part of the National Aeronautics and Space Administration's (NASA) Space Launch System (SLS) Program, engineers at NASA's Marshall Space Flight Center (MSFC) in Huntsville, Alabama are working to design, develop and implement the SLS Core Stage structural testing. The SLS will have the capability to return humans to the Moon and beyond and its first launch is scheduled for December of 2017. The SLS Core Stage consist of five major elements; Forward Skirt, Liquid Oxygen (LOX) tank, Intertank (IT), Liquid Hydrogen (LH2) tank and the Engine Section (ES). Structural Test Articles (STA) for each of these elements are being designed and produced by Boeing at Michoud Assembly Facility located in New Orleans, La. The structural test for the Core Stage STAs (LH2, LOX, IT and ES) are to be conducted by the MSFC Test Laboratory. Additionally, the MSFC Test Laboratory manages the Structural Test Equipment (STE) design and development to support the STAs. It was decided early (April 2012) in the project life that the LH2 and LOX tank STAs would require new test stands and the Engine Section and Intertank would be tested in existing facilities. This decision impacted schedules immediately because the new facilities would require Construction of Facilities (C of F) funds that require congressional approval and long lead times. The Engine Section and Intertank structural test are to be conducted in existing facilities which will limit lead times required to support the first launch of SLS. With a SLS launch date of December, 2017 Boeing had a need date for testing to be complete by September of 2017 to support flight certification requirements. The test facilities were required to be ready by October of 2016 to support test article delivery. The race was on to get the stands ready before Test Article delivery and meet the test complete date of September 2017. This paper documents the past and current design and development phases and the supporting processes, tools, and

  3. Core damage frequency (reactor design) perspectives based on IPE results

    International Nuclear Information System (INIS)

    Camp, A.L.; Dingman, S.E.; Forester, J.A.

    1996-01-01

    This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and C, modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs. Human actions found to be important in boiling water reactors (BWRs) and in pressurized water reactors (PWRs) are presented and the events most frequently found important are discussed

  4. Design and Implementation of High Level Trigger Configuration Exporter and Parser

    CERN Document Server

    Abdulwahhab, Husam

    2015-01-01

    This paper serves as a description of the project that was developed at CMS during the summer. The initial task of the project was with the design, implementation and development of a configuration exporter from an oracle database to a python file. Next was the development of a parser that reads all the necessary information from the python configuration file that was created by the parser, and store the information into the memory in the form of an efficient and easy to access and manipulate cache. The final task of the project was the implementation of a system that handles requests from the client, which is a web interface, and replies with the appropriate data organized in a way that can be viewed on the interface.

  5. Optimal design of a novel configuration of MR brake with coils placed on the side housings

    Science.gov (United States)

    Nguyen, Quoc Hung; Nguyen, Ngoc Diep; Choi, Seung-Bok

    2014-03-01

    It is well known that in design of traditional magneto-rheological brake (MRB), coils are placed on the cylindrical housing of the brake. In this study, a new configuration of MR brake with coils placed on the side housings of the brake is proposed and analyzed. After briefly explaining the operating principle of the proposed configuration, the braking torque of the MR brake is analyze based on Bingham-plastic rheological model of MR fluid. The optimization of the proposed and conventional MR brakes is then performed considering maximum braking torque and mass of the brake. Based on the optimal results, a comparison between the proposed MR brakes and the conventional ones is undertaken. In addition, experimental test of the MR brakes is conducted and the results are presented in order to validate the performance characteristics of the proposed MR brake.

  6. A supercomputing application for reactors core design and optimization

    International Nuclear Information System (INIS)

    Hourcade, Edouard; Gaudier, Fabrice; Arnaud, Gilles; Funtowiez, David; Ammar, Karim

    2010-01-01

    Advanced nuclear reactor designs are often intuition-driven processes where designers first develop or use simplified simulation tools for each physical phenomenon involved. Through the project development, complexity in each discipline increases and implementation of chaining/coupling capabilities adapted to supercomputing optimization process are often postponed to a further step so that task gets increasingly challenging. In the context of renewal in reactor designs, project of first realization are often run in parallel with advanced design although very dependant on final options. As a consequence, the development of tools to globally assess/optimize reactor core features, with the on-going design methods accuracy, is needed. This should be possible within reasonable simulation time and without advanced computer skills needed at project management scale. Also, these tools should be ready to easily cope with modeling progresses in each discipline through project life-time. An early stage development of multi-physics package adapted to supercomputing is presented. The URANIE platform, developed at CEA and based on the Data Analysis Framework ROOT, is very well adapted to this approach. It allows diversified sampling techniques (SRS, LHS, qMC), fitting tools (neuronal networks...) and optimization techniques (genetic algorithm). Also data-base management and visualization are made very easy. In this paper, we'll present the various implementing steps of this core physics tool where neutronics, thermo-hydraulics, and fuel mechanics codes are run simultaneously. A relevant example of optimization of nuclear reactor safety characteristics will be presented. Also, flexibility of URANIE tool will be illustrated with the presentation of several approaches to improve Pareto front quality. (author)

  7. Constructive and thermal design of a core fast discharge

    International Nuclear Information System (INIS)

    Schroer, H.

    1979-08-01

    The present study is concerned with the development and thermal design of a fast discharge system for balls for the PR 3000 MWsub(th) process heat reactor. The term 'fast discharge system for balls' denotes a very short-time discharge procedure of the entire core contents, i.e. the flowing out of the fuel elements due to gravity into a receiver tank underneath the prestressed-concrete vessel. From a safety-engineering point of view, the fast discharge system for balls constitutes an additional possibility of active decay heat removal, besides the multiply redundant and diversitary reactor protection system, serving to further reduce the remaining residual risk. A fast discharge system for balls, however, is to be used only in the event of all the other possibilities of active decay heat removal having failed and when the maximum permissible temperatures for particularly exposed primary circuit components have been reached. However, the application range of such a system is restricted exclusively to high-temperature reactors with spherical fuel elements; the procedure cannot be applied to other reactor systems because of the rigidly fixed position of the fuel elements inside the core and for reasons of fuel element geometry. Besides the purpose of application, the influence of in-core temperature development on the possible actuation of the fast discharge system is being described in particular detail. This is followed by a description of the structural and thermal design of three specific major components, i.e. the piping system, shut-off devices and fuel element receiver tank, which will have to be installed additionally for the implementation of a fast discharge system for balls as compared to previous plant concepts. (orig.) [de

  8. High-Speed Research: 1994 Sonic Boom Workshop. Configuration, Design, Analysis and Testing

    Science.gov (United States)

    McCurdy, David A. (Editor)

    1999-01-01

    The third High-Speed Research Sonic Boom Workshop was held at NASA Langley Research Center on June 1-3, 1994. The purpose of this workshop was to provide a forum for Government, industry, and university participants to present and discuss progress in their research. The workshop was organized into sessions dealing with atmospheric propagation; acceptability studies; and configuration design, and testing. Attendance at the workshop was by invitation only. The workshop proceedings include papers on design, analysis, and testing of low-boom high-speed civil transport configurations and experimental techniques for measuring sonic booms. Significant progress is noted in these areas in the time since the previous workshop a year earlier. The papers include preliminary results of sonic boom wind tunnel tests conducted during 1993 and 1994 on several low-boom designs. Results of a mission performance analysis of all low-boom designs are also included. Two experimental methods for measuring near-field signatures of airplanes in flight are reported.

  9. Study of advanced composite structural design concepts for an arrow wing supersonic cruise configuration, task 3

    Science.gov (United States)

    1978-01-01

    A structural design study was conducted to assess the relative merits of structural concepts using advanced composite materials for an advanced supersonic aircraft cruising at Mach 2.7. The configuration and structural arrangement developed during Task I and II of the study, was used as the baseline configuration. Allowable stresses and strains were established for boron and advanced graphite fibers based on projected fiber properties available in the next decade. Structural concepts were designed and analyzed using graphite polyimide and boron polyimide, applied to stiffened panels and conventional sandwich panels. The conventional sandwich panels were selected as the structural concept to be used on the wing structure. The upper and lower surface panels of the Task I arrow wing were redesigned using high-strength graphite polyimide sandwich panels over the titanium spars and ribs. The ATLAS computer system was used as the basis for stress analysis and resizing the surface panels using the loads from the Task II study, without adjustment for change in aeroelastic deformation. The flutter analysis indicated a decrease in the flutter speed compared to the baseline titanium wing design. The flutter analysis indicated a decrease in the flutter speed compared to the baseline titanium wing design. The flutter speed was increased to that of the titanium wing, with a weight penalty less than that of the metallic airplane.

  10. Theory of the optimal design of straight-axis minimum-B mirror confinement configurations

    International Nuclear Information System (INIS)

    Hall, L.S.

    1982-01-01

    The design of modern straight-axis linked-mirror plasma-confinement configurations involves a balance between many competing requirements. The dipole and quadrupole components of magnetic induction required in one confinement region often do not match onto the fields of an adjacent region without complications that seriously affect particle drifts or confinement stability. Here, the relevant factors are set down together with the techniques for analytical optimization of the design of a general configuration. A general sufficient condition for the stability of an arbitrary guiding-center MHD equilibrium is derived. This condition makes explicit the stabilizing qualities of good normal curvature and diamagnetic axial current. The instability drive depends on two terms: one carries the sign of normal curvature and the other relates to the relative signs of geodeics curvature and geodesic torsion. The theory is applied to low-beta, large-aspect-ratio equilibria for which analytic expressions for the confining magnetic fields are known. Two optimizations are required to specify the arbitrary features of the quadrupole and dipole fields. One optimization is nonlinear and can be performed by the ordinary calculus of variations; the second optimization is linear and subject to the rules of game theory. Appropriate quality factors are obtained, thus giving the designer quantitative measures with which to balance design trade-offs

  11. New approach to the design of core support structures for large LMFBR plants

    International Nuclear Information System (INIS)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions

  12. Tools and applications for core design and shielding in fast reactors

    International Nuclear Information System (INIS)

    Rachamin, Reuven

    2013-01-01

    Outline: • Modeling of SFR cores using the Serpent-DYN3D code sequence; • Core shielding assessment for the design of FASTEF-MYRRHA; • Neutron shielding studies on an advanced Molten Salt Fast Reactor (MSFR) design

  13. A study on the advanced statistical core thermal design methodology

    International Nuclear Information System (INIS)

    Lee, Seung Hyuk

    1992-02-01

    A statistical core thermal design methodology for generating the limit DNBR and the nominal DNBR is proposed and used in assessing the best-estimate thermal margin in a reactor core. Firstly, the Latin Hypercube Sampling Method instead of the conventional Experimental Design Technique is utilized as an input sampling method for a regression analysis to evaluate its sampling efficiency. Secondly and as a main topic, the Modified Latin Hypercube Sampling and the Hypothesis Test Statistics method is proposed as a substitute for the current statistical core thermal design method. This new methodology adopts 'a Modified Latin Hypercube Sampling Method' which uses the mean values of each interval of input variables instead of random values to avoid the extreme cases that arise in the tail areas of some parameters. Next, the independence between the input variables is verified through 'Correlation Coefficient Test' for statistical treatment of their uncertainties. And the distribution type of DNBR response is determined though 'Goodness of Fit Test'. Finally, the limit DNBR with one-sided 95% probability and 95% confidence level, DNBR 95/95 ' is estimated. The advantage of this methodology over the conventional statistical method using Response Surface and Monte Carlo simulation technique lies in its simplicity of the analysis procedure, while maintaining the same level of confidence in the limit DNBR result. This methodology is applied to the two cases of DNBR margin calculation. The first case is the application to the determination of the limit DNBR where the DNBR margin is determined by the difference between the nominal DNBR and the limit DNBR. The second case is the application to the determination of the nominal DNBR where the DNBR margin is determined by the difference between the lower limit value of the nominal DNBR and the CHF correlation limit being used. From this study, it is deduced that the proposed methodology gives a good agreement in the DNBR results

  14. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

    Armstrong, B.C.; Buechel, R.J.

    1990-01-01

    Control rod repositioning is a method for minimizing rod cluster control assembly (RCCA) wear in the upper internals area where the guide cards interface with the rodlets of the RCCAs. A number of utilities have implemented strategies for rod repositioning, which often has no impact on the nuclear analysis for cases where the control rods are never repositioned into the active fuel. Other strategies involve repositioning the control rods several steps into the active fuel. The impact of this type of repositioning on the axial power shape and consequently the total peaking factor F Q T varies, depending on the method in which the repositioning is implemented at the plant. Operating for long periods with all the control and shutdown rods inserted several steps in the active fuel followed by withdrawing them fully from the core results in a shifting of the power distribution toward the top of the core and must be accounted for in the design analysis. On the other hand, an optional plan for control rod repositioning that considers margins available in related design parameters can be devised that minimizes the effects of the repositioning for the reload. This paper summarizes a rod repositioning strategy implemented for a recent reload and some calculated power shape results for this strategy and other scenarios

  15. Simulation of Blood flow in Different Configurations Design of Bi-leaflet Mechanical Heart Valve

    Science.gov (United States)

    Hafizah Mokhtar, N.; Abas, Aizat

    2018-05-01

    In this work, two different designs of artificial heart valve were devised and then compared by considering the thrombosis, wear and valve orifice to anatomical orifice ratio of each mechanical heart valve. These different design configurations of bi-leaflet mechanical heart valves model are created through the use of Computer-aided design (CAD) modelling and simulated using Computational fluid dynamic (CFD) software. Design 1 is based on existing conventional bi-leaflet valve and design 2 based on modified bi-leaflet respectively. The flow pattern, velocity, vorticity and stress analysis have been done to justify the best design. Based on results, both of the designs show a Doppler velocity index of less than the allowable standard of 2.2 which is safe to be used as replacement of the human heart valve. However, design 2 shows that it has a lower possibility of cavitation issue which will lead to lower thrombosis and provide good central flow area of blood as compared to design 1.

  16. DENA: A Configurable Microarchitecture and Design Flow for Biomedical DNA-Based Logic Design.

    Science.gov (United States)

    Beiki, Zohre; Jahanian, Ali

    2017-10-01

    DNA is known as the building block for storing the life codes and transferring the genetic features through the generations. However, it is found that DNA strands can be used for a new type of computation that opens fascinating horizons in computational medicine. Significant contributions are addressed on design of DNA-based logic gates for medical and computational applications but there are serious challenges for designing the medium and large-scale DNA circuits. In this paper, a new microarchitecture and corresponding design flow is proposed to facilitate the design of multistage large-scale DNA logic systems. Feasibility and efficiency of the proposed microarchitecture are evaluated by implementing a full adder and, then, its cascadability is determined by implementing a multistage 8-bit adder. Simulation results show the highlight features of the proposed design style and microarchitecture in terms of the scalability, implementation cost, and signal integrity of the DNA-based logic system compared to the traditional approaches.

  17. Neutronic design of a LEU [low enriched uranium] core for the Ohio State University research reactor

    International Nuclear Information System (INIS)

    Seshadri, M.D.; Aybar, H.S.; Aldemir, T.

    1987-01-01

    The 10 kw HEU fuelled Ohio State University Reactor (OSURR) will be upgraded to operate at 500 kW with standardized 125 g 235 U LEU U 3 Si 2 fuel plates. An earlier scoping study based on two-dimensional diffusion calculations has identified the potential LEU core configurations for the conversion/upgrade of OSURR using the standardized plates in a 16-plate (+ 2 dummy plates) standard and 10-scoping study is improved for a more precise determination of the excess reactivities and safety rod worths for these potential configurations. Comparison of the results obtained by the improved model to experimental results and to the results of full-core Monte Carlo simulations shows excellent agreement. The results also indicate that the conversion/upgrade of OSURR can be realized with three possible LEU core configurations while maintaining a cold, clean shutdown margin of 1.57-1.91 % Δ k/k, depending on the configuration used. (Author)

  18. Seismic responses of a pool-type fast reactor with different core support designs

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

  19. Core/shell PLGA microspheres with controllable in vivo release profile via rational core phase design.

    Science.gov (United States)

    Yu, Meiling; Yao, Qing; Zhang, Yan; Chen, Huilin; He, Haibing; Zhang, Yu; Yin, Tian; Tang, Xing; Xu, Hui

    2018-02-27

    the microspheres prepared by various methods were mainly controlled by either the porosity inside the microspheres or the degradation of materials, which could, therefore, lead to different release behaviours. This results indicated great potential of the PLGA microsphere formulation as an injectable depot for controllable in vivo release profile via rational core phase design. Core/shell microspheres fabricated by modified double emulsification-solvent evaporation methods, with various inner phases, to obtain high loading drugs system, as well as appropriate release behaviours. Accordingly, control in vivo release profile via rational core phase design.

  20. Dynamic behavior of homogeneous and heterogeneous LMFBR core-design concepts

    International Nuclear Information System (INIS)

    Chang, Y.I.; Henryson, H. II; Orechwa, Y.; Su, S.F.; Greenman, G.; Blomquist, R.

    1981-01-01

    The emphasis is placed on obtaining an understanding of the inherent difference between homogeneous and heterogeneous core configurations regarding neutronic characteristics related to the dynamic behavior. The space-time neutronic and thermal-hydraulic behavior was analyzed in detail for various core configurations by using the FX2-TH, a two-dimensional kinetics code with thermal-hydraulic feedback. In addition, the relationship between the flux tilt and the fundamental-to-first harmonic eigenvalue separation, and the sodium void reactivity in heterogeneous cores were also sutdied

  1. Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

    International Nuclear Information System (INIS)

    Lim, Hong Sik; Kim, Min Hwan

    2009-01-01

    The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

  2. Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor

    Directory of Open Access Journals (Sweden)

    Bong Guen Hong

    2018-02-01

    Full Text Available A configuration of a fusion-driven transmutation reactor with a low aspect ratio tokamak-type neutron source was determined in a self-consistent manner by using coupled analysis of tokamak systems and neutron transport. We investigated the impact of blanket configuration on the characteristics of a fusion-driven transmutation reactor. It was shown that by merging the TRU burning blanket and tritium breeding blanket, which uses PbLi as the tritium breeding material and as coolant, effective transmutation is possible. The TRU transmutation capability can be improved with a reduced blanket thickness, and fast fluence at the first wall can be reduced.  Article History: Received: July 10th 2017; Received: Dec 17th 2017; Accepted: February 2nd 2018; Available online How to Cite This Article: Hong, B.G. (2018 Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor. International Journal of Renewable Energy Development, 7(1, 65-70. https://doi.org/10.14710/ijred.7.1.65-70

  3. The space shuttle ascent vehicle aerodynamic challenges configuration design and data base development

    Science.gov (United States)

    Dill, C. C.; Young, J. C.; Roberts, B. B.; Craig, M. K.; Hamilton, J. T.; Boyle, W. W.

    1985-01-01

    The phase B Space Shuttle systems definition studies resulted in a generic configuration consisting of a delta wing orbiter, and two solid rocket boosters (SRB) attached to an external fuel tank (ET). The initial challenge facing the aerodynamic community was aerodynamically optimizing, within limits, this configuration. As the Shuttle program developed and the sensitivities of the vehicle to aerodynamics were better understood the requirements of the aerodynamic data base grew. Adequately characterizing the vehicle to support the various design studies exploded the size of the data base to proportions that created a data modeling/management challenge for the aerodynamicist. The ascent aerodynamic data base originated primarily from wind tunnel test results. The complexity of the configuration rendered conventional analytic methods of little use. Initial wind tunnel tests provided results which included undesirable effects from model support tructure, inadequate element proximity, and inadequate plume simulation. The challenge to improve the quality of test results by determining the extent of these undesirable effects and subsequently develop testing techniques to eliminate them was imposed on the aerodynamic community. The challenges to the ascent aerodynamics community documented are unique due to the aerodynamic complexity of the Shuttle launch. Never before was such a complex vehicle aerodynamically characterized. The challenges were met with innovative engineering analyses/methodology development and wind tunnel testing techniques.

  4. A novel design of reactive distillation configuration for 2-methoxy-2-methylheptane process

    Directory of Open Access Journals (Sweden)

    Hussain Arif

    2017-01-01

    Full Text Available The study aims to reveal the possibility of reactive distillation (RD in the 2-methoxy-2-methylheptane (MMH production process. MMH is getting more industrial and academic interests as a gasoline additive to replace methyl tert-butyl ether. Traditionally, MMH is obtained by carrying out the reaction in the reactor followed by three distillation columns. The high yield of MMH could be achieved by keeping the large reactor size or by using the large excess of 2-methyl-1-heptene (MH. Both former and latter strategies are associated with the high capital and operating costs. To solve these problems, this study proposed an innovative RD configuration to take synergistic benefits of reaction and separation involved. This innovative RD configuration allows the production of MMH with significantly lower capital, operating and total annual costs. For desired MMH yield, the result demonstrates that the proposed RD configuration can reduce energy, capital, and total annual costs up to 7.7, 31.3, and 17.1%, respectively, compared to a conventional process. Furthermore, the influence of some important design parameters on the RD column performance was also explored to overcome the temperature limitation of acid resin catalyst inside the reactive zone of the RD column.

  5. A new divertor plates design concept for the double null NET configuration

    International Nuclear Information System (INIS)

    Farfaletti-Casali, F.; Renda, V.; Federici, G.; Papa, L.

    1986-01-01

    A new divertor plate design concept for the Double Null NET configuration (NET-DN) is presented. This concept applies to the plasma configuration of NET and takes advantage by the maintenance scheme of the internal components adopted in NET. According to this maintenance approach, which uses the top loading of the internal segments, 48 inboard removable segments, 3 for each of the 16 reactor sectors, act as simple protective panels, gathering together in only one piece the plates of both the upper and lower divertor regions and the intermediate portion of the inboard first wall. They are cooled by water flowing inside a set of hairpin-shaped, stainless steel tubes, arranged in poloidal direction inside a copper heat sink, and fed by supply lines at the top of the reactor. The surface facing the plasma is covered by a tungsten alloy layer. In such a way, the maintenance of the two divertor regions and of the inboard first wall can be easily achieved by removing the inboard panels from the top of the reactor. The layout of the cooling system and preliminary thermohydraulics and thermomechanical calculations, carried out for assessing the feasibility of the proposed system for the NET reference configuration, are reported in this paper. (author)

  6. A new divertor plates design concept for the double null net configuration

    International Nuclear Information System (INIS)

    Farfaletti-Casali, F.; Iop, O.; Renda, V.; Federici, G.; Papa, L.

    1987-01-01

    A new divertor plate design concept for the Double Null NET configuration (NET-DN) is presented in this paper. This concept applies to the plasma configuration of NET and takes advantage by the maintenance scheme of the internal components adopted in NET. According to this maintenance approach, which uses the top loading of the internal segments, 48 inboard removable segments, 3 for each of the 16 reactor sectors, act as simple protective panels, gathering together in only one piece the plates of both the upper and lower divertor regions and the intermediate portion of the inboard first wall. They are cooled by water flowing inside a set of hairpin-shaped, stainless steel tubes, arranged in poloidal direction inside a copper heat sink, and fed by supply lines at the top of the reactor. The surface facing the plasma is covered by a tungsten alloy layer. In such a way, the maintenance of the two divertor regions and of the inboard first wall can be easily achieved by removing the inboard panels from the top of the reactor. The layout of the cooling system and preliminary thermohydraulics and thermomechanical calculations, carried out for assessing the feasibility of the proposed system for the NET reference configuration, are reported in this paper

  7. Ergonomic design of mosaic control panel and standardised control tile configurations for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Ughade, A.V.; Das, R.N.; Ramakrishnan, S.

    1994-01-01

    A review of control rooms of operating nuclear power plants identified many design problems having potential for degrading the performance of operators. Many indications and controls on existing control panels are placed outside the recommended visual and reach envelopes for acceptable operator usage. As a result, the application of human factor principles was found to be needed. This paper describes the design approach for working out the dimensions of main control room panels and console using human engineering principles and recommends the ergonomic dimensions of the main control room panels and console. Further it gives the basis and works out the control tile configurations for 500 MWe PHWR project. It also suggests the use of a full scale mock up for design evaluation and verification. (author). 7 refs., 4 figs

  8. Parametric Flutter Analysis of the TCA Configuration and Recommendation for FFM Design and Scaling

    Science.gov (United States)

    Baker, Myles; Lenkey, Peter

    1997-01-01

    The current HSR Aeroelasticity plan to design, build, and test a full span, free flying transonic flutter model in the TDT has many technical obstacles that must be overcome for a successful program. One technical obstacle is the determination of a suitable configuration and point in the sky to use in setting the scaling point for the ASE models program. Determining this configuration and point in the sky requires balancing several conflicting requirements, including model buildability, tunnel test safety, and the ability of the model to represent the flutter mechanisms of interest. As will be discussed in detail in subsequent sections, the current TCA design exhibits several flutter mechanisms of interest. It has been decided that the ASE models program will focus on the low frequency symmetric flutter mechanism, and will make no attempt to investigate high frequency flutter mechanisms. There are several reasons for this choice. First, it is believed that the high frequency flutter mechanisms are similar in nature to classical wing bending/torsion flutter, and therefore there is more confidence that this mechanism can be predicted using current techniques. The low frequency mode, on the other hand, is a highly coupled mechanism involving wing, body, tail, and engine motion which may be very difficult to predict. Second, the high frequency flutter modes result in very small weight penalties (several hundred pounds), while suppression of the low frequency mechanism inside the flight envelope causes thousands of pounds to be added to the structure. In order to successfully test the low frequency flutter mode of interest, a suitable starting configuration and point in the sky must be identified. The configuration and point in the sky must result in a wind tunnel model that (1) represents the low-frequency wing/body/engine/empennage flutter mechanisms that are unique to HSCT configurations, (2) flutters at an acceptably low frequency in the tunnel, (3) flutters at an

  9. ORR core re-configuration measurements to increase the fast neutron flux in the Magnetic Fusion Energy (MFE) experiments

    International Nuclear Information System (INIS)

    Hobbs, R.W.; Stinnett, R.M.; Sims, T.M.

    1985-06-01

    A study has been made of the relative increases obtainable in the fast neutron flux in the Magnetic Fusion Energy (MFE) experiment positions by reconfiguring the current ORR core. The study was made at the request of the MFE program to examine the percentage increase possible in the current displacement per atom (dpa) rate (assumed proportional to the fast flux). The principle methods investigated to increase the fast flux consisted of reducing the current core size (number of fuel elements) to increase the core average power density and arrangement of the fuel elements in the reduced-size core to tilt the core power distribution towards the MFE positions. The study concluded that fast fluxes in the E-3 core position could be increased by approximately 15 to 20% over current values and in E-5 by approximately 45 to 55%

  10. Exploring Many-Core Design Templates for FPGAs and ASICs

    Directory of Open Access Journals (Sweden)

    Ilia Lebedev

    2012-01-01

    Full Text Available We present a highly productive approach to hardware design based on a many-core microarchitectural template used to implement compute-bound applications expressed in a high-level data-parallel language such as OpenCL. The template is customized on a per-application basis via a range of high-level parameters such as the interconnect topology or processing element architecture. The key benefits of this approach are that it (i allows programmers to express parallelism through an API defined in a high-level programming language, (ii supports coarse-grained multithreading and fine-grained threading while permitting bit-level resource control, and (iii reduces the effort required to repurpose the system for different algorithms or different applications. We compare template-driven design to both full-custom and programmable approaches by studying implementations of a compute-bound data-parallel Bayesian graph inference algorithm across several candidate platforms. Specifically, we examine a range of template-based implementations on both FPGA and ASIC platforms and compare each against full custom designs. Throughout this study, we use a general-purpose graphics processing unit (GPGPU implementation as a performance and area baseline. We show that our approach, similar in productivity to programmable approaches such as GPGPU applications, yields implementations with performance approaching that of full-custom designs on both FPGA and ASIC platforms.

  11. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

  12. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    International Nuclear Information System (INIS)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun

    2016-01-01

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes

  13. Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

    International Nuclear Information System (INIS)

    Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

    2008-01-01

    In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

  14. A multi-perspective approach for the design of Product Configuration Systems - an evaluation of industry applications

    DEFF Research Database (Denmark)

    Hvam, Lars

    2004-01-01

    , a de-marcation and definition of the configuration system to be designed. • Analysis and modelling of the part of the company’s product assortment which is to be included in the con-figuration system. • Selection of configuration software and programming of the configuration system. • Implementation....... The procedure or certain parts of the procedure have currently been tested and further developed in cooperation with a number of industrial companies including F.L.Smidth, American Power Conversion (APC), Aalborg industries, NEG-Micon, GEA-Niro and IBM-SMS. This paper presents the experiences gained from 4...

  15. Business model design through a designer's lens : Translating, transferring and transforming cognitive configurations into action

    NARCIS (Netherlands)

    Simonse, W.L.; Badke-Schaub, P.G.

    2015-01-01

    Strategic managers are challenged to take advantage of digitalisation opportunities related to services of social media and web 2.0 technologies. Business innovations such as crowd sourcing platforms require a new way of integrating business to technology, articulated in a new business model designs

  16. Design and Implementation of Cloud-Centric Configuration Repository for DIY IoT Applications.

    Science.gov (United States)

    Ahmad, Shabir; Hang, Lei; Kim, Do Hyeun

    2018-02-06

    The Do-It-Yourself (DIY) vision for the design of a smart and customizable IoT application demands the involvement of the general public in its development process. The general public lacks the technical knowledge for programming state-of-the-art prototyping and development kits. The latest IoT kits, for example, Raspberry Pi, are revolutionizing the DIY paradigm for IoT, and more than ever, a DIY intuitive programming interface is required to enable the masses to interact with and customize the behavior of remote IoT devices on the Internet. However, in most cases, these DIY toolkits store the resultant configuration data in local storage and, thus, cannot be accessed remotely. This paper presents the novel implementation of such a system, which not only enables the general public to customize the behavior of remote IoT devices through a visual interface, but also makes the configuration available everywhere and anytime by leveraging the power of cloud-based platforms. The interface enables the visualization of the resources exposed by remote embedded resources in the form of graphical virtual objects (VOs). These VOs are used to create the service design through simple operations like drag-and-drop and the setting of properties. The configuration created as a result is maintained as an XML document, which is ingested by the cloud platform, thus making it available to be used anywhere. We use the HTTP approach for the communication between the cloud and IoT toolbox and the cloud and real devices, but for communication between the toolbox and actual resources, CoAP is used. Finally, a smart home case study has been implemented and presented in order to assess the effectiveness of the proposed work.

  17. Design and Implementation of Cloud-Centric Configuration Repository for DIY IoT Applications

    Directory of Open Access Journals (Sweden)

    Shabir Ahmad

    2018-02-01

    Full Text Available The Do-It-Yourself (DIY vision for the design of a smart and customizable IoT application demands the involvement of the general public in its development process. The general public lacks the technical knowledge for programming state-of-the-art prototyping and development kits. The latest IoT kits, for example, Raspberry Pi, are revolutionizing the DIY paradigm for IoT, and more than ever, a DIY intuitive programming interface is required to enable the masses to interact with and customize the behavior of remote IoT devices on the Internet. However, in most cases, these DIY toolkits store the resultant configuration data in local storage and, thus, cannot be accessed remotely. This paper presents the novel implementation of such a system, which not only enables the general public to customize the behavior of remote IoT devices through a visual interface, but also makes the configuration available everywhere and anytime by leveraging the power of cloud-based platforms. The interface enables the visualization of the resources exposed by remote embedded resources in the form of graphical virtual objects (VOs. These VOs are used to create the service design through simple operations like drag-and-drop and the setting of properties. The configuration created as a result is maintained as an XML document, which is ingested by the cloud platform, thus making it available to be used anywhere. We use the HTTP approach for the communication between the cloud and IoT toolbox and the cloud and real devices, but for communication between the toolbox and actual resources, CoAP is used. Finally, a smart home case study has been implemented and presented in order to assess the effectiveness of the proposed work.

  18. Ideas to Design an in situ Diamond Drilling Core Splitter within Soft ...

    African Journals Online (AJOL)

    Michael O. Mensah

    2015-12-02

    Dec 2, 2015 ... the wireline system of core barrel assembly and the device used in splitting of core ... Keywords: Design, In situ, Diamond drilling, Core splitter, Wireline system .... This is the most complex part of the core barrel and has many.

  19. Fast reactor calculational route for Pu burning core design

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  20. Design studies of back up cores for the experimental multi-purpose VHTR, (1)

    International Nuclear Information System (INIS)

    Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu

    1982-09-01

    For the Experimental Multi-Purpose Very High Temperature Reactor, design studies have been made of two backup cores loaded with new type fuel elements. The purpose is to improve core operational characteristics of the standard design core (Mark-III core) consisting of pin-in-block type fuel element having externally cooled hollow fuel rods. The first backup core (semi-pin fuel core) is composed of fuel elements with internally cooled fuel pins, and the second core (multihole fuel core) is composed of multihole fuel elements, which can be adopted for the experimental VHTR as the substitution of the standard Mark-III fuel element. Either of the cores has 73 fuel columns and 4 m height. The arrangement of active core and reactor internal structure is same as that in the standard design core. These backup cores meet almost all design requirements of the VHTR and increase the margins for some important design items in comparison with the standard core (Mark-III core). This report describes the overall characteristics of nuclear, thermal-hydraulic, fuel and safety, and structural consideration for these cores. (author)

  1. Using Computational Fluid Dynamics and Experiments to Design Sweeping Jets for High Reynolds Number Cruise Configurations

    Science.gov (United States)

    Jones, Gregory S.; Milholen, William E., II; Fell, Jared S.; Webb, Sandy R.; Cagle, C. Mark

    2016-01-01

    The application of a sweeping jet actuator to a circulation control system was initiated by a risk reduction series of experiments to optimize the authority of a single sweeping jet actuator. The sweeping jet design was integrated into the existing Fundamental Aerodynamic Subsonic Transonic- Modular Active Control (FAST-MAC) model by replacing the steady blowing system with an array of thirty-nine sweeping jet cartridges. A constant slot height to wing chord ratio was similar to the steady blowing configuration resulting in each actuator having a unique in size for the sweeping jet configuration. While this paper will describe the scaling and optimization of the actuators for future high Reynolds number applications, the major focus of this effort was to target the transonic flight regime by increasing the amplitude authority of the actuator. This was accomplished by modifying the diffuser of the sweeping jet actuator, and this paper highlights twelve different diffuser designs. The experimental portion of this work was completed in the NASA Langley National Transonic Facility.

  2. Structural design of the toroidal configuration of the HTS SMES cooling system

    International Nuclear Information System (INIS)

    Yeom, H.K.; Koh, D.Y.; Ko, J.S.; Kim, H.B.; Hong, Y.J.; Kim, S.H.; Seong, K.C.

    2011-01-01

    The superconducting magnetic energy storage (SMES) system is working on around 30 K, because the magnet is made of high temperature superconductor. To maintain the cryogenic temperature, the superconducting coil is cooled by cryogen, helium gas or liquid neon. But there are some weak points in the cryogen cooling system. For example periodic charge of the cryogen and size is big and so on. So, we have designed the conduction cooling system for toroidal configuration HTS SMES. The toroidal type HTS SMES has some merits, so it is very small magnetic field leakage, and magnetic field applied perpendicular to the tape surface can be reduced. Our system has 28 numbers of HTS double pancake coils and they are arrayed toroidal configuration. The toroidal inner radius is 162 mm, and outer radius is 599 mm, and height is about 162 mm. In this study, we have designed the cooling structure and analyzed temperature distribution of cooling path, thermal stress and deformation of the cooling structure.

  3. Conceptual design study of LMFBR core with carbide fuel

    International Nuclear Information System (INIS)

    Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

    1987-01-01

    Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

  4. Alternative Shear Panel Configurations for Light Wood Construction. Development, Seismic Performance, and Design Guidance

    Science.gov (United States)

    Wilcoski, James; Fischer, Chad; Allison, Tim; Malach, Kelly Jo

    2002-04-01

    Shear panels are used in light wood construction to resist lateral loads resulting from earthquakes or strong winds. These panels are typically made of wooden sheathing nailed to building frame members, but this standard panel design interferes with the installation of sheet insulation. A non-insulated shear panel conducts heat between the building interior and exterior wasting considerable amounts of energy. Several alternative shear panel designs were developed to avoid this insulation-mounting problem and sample panels were tested according to standard cyclic test protocols. One of the alternative designs consisted of diagonal steel straps nailed directly to the structural framing. Several others consisted of sheathing nailed to 2 x 4 framing then set into a larger 2 x 6 structural frame in such a way that no sheathing protruded beyond the edge of the 2 x 6 members. Also samples of industry-standard shear panels were constructed and tested in order to establish a performance baseline. Analytical models were developed to size test panels and predict panel behavior. A procedure was developed for establishing design capacities based on both test data and established baseline panel design capacity. The behavior of each panel configuration is documented and recommended design capacities are presented.

  5. Design Core Commonalities: A Study of the College of Design at Iowa State University

    Science.gov (United States)

    Venes, Jane

    2015-01-01

    This comprehensive study asks what a group of rather diverse disciplines have in common. It involves a cross-disciplinary examination of an entire college, the College of Design at Iowa State University. This research was intended to provide a sense of direction in developing and assessing possible core content. The reasoning was that material…

  6. Development of UCMS for Analysis of Designed and Measured Core Power Distribution

    International Nuclear Information System (INIS)

    Moon, Sang Rae; Hong, Sun Kwan; Yang, Sung Tae

    2009-01-01

    In this study, reactor core loading patterns were determined by calculating and verifying the factors affecting peak power and important core safety variables were reconciled with their design criteria using a newly designed unified core management system. Core loading patterns are designed for quadrant cores under the assumption that the power distribution of the reactor core is the same among symmetric fuel assemblies within the core. Actual core power distributions measured during core operation may differ slightly from their designed data. Reactor engineers monitor these differences between the designed and measured data by performing a surveillance procedure every month according to the technical specification requirements. It is difficult to monitor overall power distribution behavior throughout the assemblies using the current procedure because it requires the reactor engineer to compare the designed data with only the maximum value of the power peaking factor and the relative power density. It is necessary to enhance this procedure to check the primary variables such as core power distribution, because long cycle operation, high burnup, power up-rate, and improved fuel can change the environment in the core. To achieve this goal, a web-based Unified Core Management System (UCMS) was developed. To build the UCMS, a database system was established using reactor design data such as that in the Nuclear Design Report (NDR) and automated core analysis codes for all light water reactor power plants. The UCMS is designed to help reactor engineers to monitor important core variables and core safety margins by comparing the measured core power distribution with designed data for each fuel assembly during the cycle operation in nuclear power plants

  7. 78 FR 32988 - Core Principles and Other Requirements for Designated Contract Markets; Correction

    Science.gov (United States)

    2013-06-03

    ... COMMODITY FUTURES TRADING COMMISSION 17 CFR Part 38 RIN 3038-AD09 Core Principles and Other... regarding Core Principles and Other Requirements for Designated Contract Markets by inserting a missing... regarding Core Principles and Other Requirements for Designated Contract Markets (77 FR 36612, June 19, 2012...

  8. Design of stationary PEFC system configurations to meet heat and power demands

    Science.gov (United States)

    Wallmark, Cecilia; Alvfors, Per

    This paper presents heat and power efficiencies of a modeled PEFC system and the methods used to create the system configuration. The paper also includes an example of a simulated fuel cell system supplying a building in Sweden with heat and power. The main method used to create an applicable fuel cell system configuration is pinch technology. This technology is used to evaluate and design a heat exchanger network for a PEFC system working under stationary conditions, in order to find a solution with high heat utilization. The heat exchanger network in the system connecting the reformer, the burner, gas cleaning, hot-water storage and the PEFC stack will affect the heat transferred to the hot-water storage and thereby the heating of the building. The fuel, natural gas, is reformed to a hydrogen-rich gas within a slightly pressurized system. The fuel processor investigated is steam reforming, followed by high- and low-temperature shift reactors and preferential oxidation. The system is connected to the electrical grid for backup and peak demands and to a hot-water storage to meet the varying heat demand for the building. The procedure for designing the fuel cell system installation as co-generation system is described, and the system is simulated for a specific building in Sweden during 1 year. The results show that the fuel cell system in combination with a burner and hot-water storage could supply the building with the required heat without exceeding any of the given limitations. The designed co-generation system will provide the building with most of its power requirements and would further generate income by sale of electricity to the power grid.

  9. Back up core designs for the experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Aochi, Tetsuo; Yasuno, Takehiko; Miyamoto, Yoshiaki; Shindo, Ryuichi; Ikushima, Takeshi

    1979-02-01

    For the Experimental Multi-Purpose Very High Temperature Reactor (thermal power 50 MW and reactor outlet helium temperature 1000 0 C), design studies have been made of two backup cores loaded with new-type fuel elements. The purpose is to improve core operational characteristics, especially in thermohydraulics, of the reference design core consisting of pin-in-block type fuel elements having externally cooled hollow fuel rods. In this report are described the design principles and the analyses made of nuclear, thermal and hydraulic, fuel, and safety performances to determine the backup fuel and core design parameters. The first backup core (SP fuel core) is composed of fuel elements with internally cooled fuel rods (semi-pin), 36 rods in each standard element and 18 rods in each control element. The second backup core (MH fuel core) is composed of multihole fuel elements. 102 fuel and 54 coolant holes in each standard element and 30 fuel and 18 coolant holes in each control element. Either of the cores has 73 fuel columns 4 m high; the arrangement of active core and reactor internal structures is the same as that in the reference design. The backup cores meet nearly all design requirements of the VHTR, permitting the rated power operation with coolant Reynolds number of over 10,000 in the SP core and over 6,000 in the MH core. (author)

  10. Hypersonic Separated Flows About "Tick" Configurations With Sensitivity to Model Design

    Science.gov (United States)

    Moss, J. N.; O'Byrne, S.; Gai, S. L.

    2014-01-01

    This paper presents computational results obtained by applying the direct simulation Monte Carlo (DSMC) method for hypersonic nonequilibrium flow about "tick-shaped" model configurations. These test models produces a complex flow where the nonequilibrium and rarefied aspects of the flow are initially enhanced as the flow passes over an expansion surface, and then the flow encounters a compression surface that can induce flow separation. The resulting flow is such that meaningful numerical simulations must have the capability to account for a significant range of rarefaction effects; hence the application of the DSMC method in the current study as the flow spans several flow regimes, including transitional, slip, and continuum. The current focus is to examine the sensitivity of both the model surface response (heating, friction and pressure) and flowfield structure to assumptions regarding surface boundary conditions and more extensively the impact of model design as influenced by leading edge configuration as well as the geometrical features of the expansion and compression surfaces. Numerical results indicate a strong sensitivity to both the extent of the leading edge sharpness and the magnitude of the leading edge bevel angle. Also, the length of the expansion surface for a fixed compression surface has a significant impact on the extent of separated flow.

  11. The impact of the operating environment on the design of redundant configurations

    International Nuclear Information System (INIS)

    Marseguerra, M.; Padovani, E.; Zio, E.

    1999-01-01

    Safety systems are often characterized by substantial redundancy and diversification in safety critical components. In principle, such redundancy and diversification can bring benefits when compared to single-component systems. However, it has also been recognized that the evaluation of these benefits should take into account that redundancies cannot be founded, in practice, on the assumption of complete independence, so that the resulting risk profile is strongly dominated by dependent failures. It is therefore mandatory that the effects of common cause failures be estimated in any probabilistic safety assessment (PSA). Recently, in the Hughes model for hardware failures and in the Eckhardt and Lee models for software failures, it was proposed that the stressfulness of the operating environment affects the probability that a particular type of component will fail. Thus, dependence of component failure behaviors can arise indirectly through the variability of the environment which can directly affect the success of a redundant configuration. In this paper we investigate the impact of indirect component dependence by means of the introduction of a probability distribution which describes the variability of the environment. We show that the variance of the distribution of the number, or times, of system failures can give an indication of the presence of the environment. Further, the impact of the environment is shown to affect the reliability and the design of redundant configurations

  12. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  13. Rapidly Re-Configurable Flight Simulator Tools for Crew Vehicle Integration Research and Design

    Science.gov (United States)

    Pritchett, Amy R.

    2002-01-01

    While simulation is a valuable research and design tool, the time and difficulty required to create new simulations (or re-use existing simulations) often limits their application. This report describes the design of the software architecture for the Reconfigurable Flight Simulator (RFS), which provides a robust simulation framework that allows the simulator to fulfill multiple research and development goals. The core of the architecture provides the interface standards for simulation components, registers and initializes components, and handles the communication between simulation components. The simulation components are each a pre-compiled library 'plugin' module. This modularity allows independent development and sharing of individual simulation components. Additional interfaces can be provided through the use of Object Data/Method Extensions (OD/ME). RFS provides a programmable run-time environment for real-time access and manipulation, and has networking capabilities using the High Level Architecture (HLA).

  14. Analytic function expansion nodal method for nuclear reactor core design

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

    than the analytic function. The second variation of the AFEN method we developed is the AFEN/PEN hybrid method. This method is designed especially for the multigroup reactor analysis. This hybrid method solves the diffusion equations for the fast energy groups by the PEN method, and those for the thermal energy groups by the AFEN method. This method is based on the observation that the fast group neutron flux distributions are generally so smooth that they can be approximated by a high-order polynomial and that, on the other hand, the thermal fluxes require the analytic function expansion for the representation of their strong gradients near the interface between assemblies having different neutronic properties. The results of benchmark problems on which this method was tested indicate that performance of the hybrid method is much better than that of the PEN method and is nearly the same to that of the AFEN method. In order for the AFEN method and its variations to be used in analyzing the neutron behavior in an actual reactor core, we also developed a new burnup correction model to reduce the errors in nodal flux distributions induced by the intranodal burnup gradients. It is essential for the nodal methods to maintain their accuracy in fuel depletion analysis. The burnup correction model developed in this study homogenizes equivalently the node with the burnup-induced cross section variations into the homogeneous node with the equivalent parameters such as the flux-volume-weighted constant cross sections and the discontinuity factors. The results of a benchmark problem show that this model eliminates almost all the errors in the nodal unknowns which are induced by the intranodal burnup gradients

  15. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  16. Core design for use with precision composite reflectors

    Science.gov (United States)

    Porter, Christopher C. (Inventor); Jacoy, Paul J. (Inventor); Schmitigal, Wesley P. (Inventor)

    1992-01-01

    A uniformly flexible core, and method for manufacturing the same, is disclosed for use between the face plates of a sandwich structure. The core is made of a plurality of thin corrugated strips, the corrugations being defined by a plurality of peaks and valleys connected to one another by a plurality of diagonal risers. The corrugated strips are orthogonally criss-crossed to form the core. The core is particularly suitable for use with high accuracy spherically curved sandwich structures because undesirable stresses in the curved face plates are minimized due to the uniform flexibility characteristics of the core in both the X and Y directions. The core is self venting because of the open geometry of the corrugations. The core can be made from any suitable composite, metal, or polymer. Thermal expansion problems in sandwich structures may be minimized by making the core from the same composite materials that are selected in the manufacture of the curved face plates because of their low coefficients of thermal expansion. Where the strips are made of a composite material, the core may be constructed by first cutting an already cured corrugated sheet into a plurality of corrugated strips and then secondarily bonding the strips to one another or, alternatively, by lying a plurality of uncured strips orthogonally over one another in a suitable jig and then curing and bonding the entire plurality of strips to one another in a single operation.

  17. Study on core flow distribution of the reference core design Mark-III of experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Satoh, Sadao; Arai, Taketoshi; Miyamoto, Yoshiaki; Hirano, Mitsumasa

    1977-01-01

    Concerning the coolant flow distribution between fuel channels and other flow paths in the core, designated as Reference Core Mark-III of the Multi-purpose Experimental Very High Temperature Reactor, thermal analysis has been made of the control rods and other steel structures around the core to find the coolant flow rates (bypass flow) necessary to cool them to their safe operating temperatures. Calculations showed that adequate cooling could be achieved in the Mark-III Core by the bypass flow of 8% of the total reactor coolant flow, 4% each for the control-rod channels and for other structures. The thermal and coolant flow design bases, including the assumption of a 10% bypass flow, were thus confirmed to first approximation. (auth.)

  18. A latency analysis for M2M and OG-like traffic patterns in different HSPA core network configurations

    Directory of Open Access Journals (Sweden)

    M. V. Popović

    2014-11-01

    Full Text Available In this paper we present an analysis intended to reveal possible impacts of core network features on latency for modelled M2M and Online Gaming traffic. Simulations were performed in a live 3G/HSPA network. Test traffic simulating multiplayer real-time games and M2M applications was generated on 10 mobile phones in parallel, sending data to a remote server. APNs with different combinations of hardware and features (proxy server, different GGSNs and firewalls, usage of Service Awareness feature were chosen. The traffic was recorded on the Gn interface in the mobile core. The goal of experiments was to evaluate any eventually significant variation of average recorded RTTs in the core part of mobile network that would clearly indicate either the impact of used APN on delay for a specific traffic pattern, or selectivity of the APN towards different traffic patterns.

  19. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    plenum and boron layer, varying the core height-to-diameter (H/D) ratio, introducing moderator pins into the fuel assemblies, and modifying the initially loaded plutonium content. The sensitivity of the principal safety and performance parameters, viz. void reactivity, Doppler constant, nominal reactivity and breeding gain, has been evaluated with respect to each of the options. Finally, two synthesis core concepts, representing different selected combinations of the optimization options, have been proposed. The third and last phase has been to test the proposed optimized designs in terms of the dynamic core response to a representative ULOF accident, in which sodium boiling occurs. Such a hypothetical transient is of prime importance for SFR safety demonstration, since it may lead to a severe accident situation, where both cladding and fuel melt. As compared to the response of the reference ESFR core, certain improvements are indeed achieved with the static neutronics optimization carried out. However, these are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting. Thermal-hydraulic optimization has thus been considered necessary to: 1) prevent sodium flow blockage in the fuel channels and 2) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. Following implementation of appropriate thermal-hydraulics related design changes, one arrives at a final configuration of the SFR core in which, for the selected accident scenario, a new ‘steady state’ involving stable sodium boiling is shown to be achievable, with melting of neither cladding nor fuel. Such satisfactory behavior has been confirmed not only for the beginning-of-life core state, but also for the equilibrium closed fuel cycle. (author)

  20. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    and boron layer, varying the core height-to-diameter (H/D) ratio, introducing moderator pins into the fuel assemblies, and modifying the initially loaded plutonium content. The sensitivity of the principal safety and performance parameters, viz. void reactivity, Doppler constant, nominal reactivity and breeding gain, has been evaluated with respect to each of the options. Finally, two synthesis core concepts, representing different selected combinations of the optimization options, have been proposed. The third and last phase has been to test the proposed optimized designs in terms of the dynamic core response to a representative ULOF accident, in which sodium boiling occurs. Such a hypothetical transient is of prime importance for SFR safety demonstration, since it may lead to a severe accident situation, where both cladding and fuel melt. As compared to the response of the reference ESFR core, certain improvements are indeed achieved with the static neutronics optimization carried out. However, these are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting. Thermal-hydraulic optimization has thus been considered necessary to: 1) prevent sodium flow blockage in the fuel channels and 2) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. Following implementation of appropriate thermal-hydraulics related design changes, one arrives at a final configuration of the SFR core in which, for the selected accident scenario, a new ‘steady state’ involving stable sodium boiling is shown to be achievable, with melting of neither cladding nor fuel. Such satisfactory behavior has been confirmed not only for the beginning-of-life core state, but also for the equilibrium closed fuel cycle. (author)

  1. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  2. Neutronic design of a traveling wave reactor core

    International Nuclear Information System (INIS)

    Lopez S, R. C.; Francois L, J. L.

    2010-10-01

    The traveling wave reactor is an innovative kind of fast breeder reactor, capable of operate for decades without refueling and whose operation requires only a small amount of enriched fuel for the ignition. Also, one of its advantages is its versatility; it can be designed as small modules of about 100 M We or large scale units of 1000 M We. In this paper the behaviour of the traveling wave reactor core is studied in order to determine whether the traveling breeding/burning wave moves (as theoretically predicted) or not. To achieve this, we consider a two pieces cylinder, the first one, the ignition zone, containing highly enriched fuel and the second, the breeding zone, which is the larger, containing natural or depleted uranium or thorium. We consider that both zones are homogeneous mixtures of fuel, sodium as coolant and iron as structural material. We also include a reflector material outside the cylinder to reduce the neutron leakages. Simulations were run with MCNPX version 2.6 code. We observed that the wave does move as time passes as predicted by theory, and reactor remains supercritical in the time we have simulated (3000 days). Also, we found that thorium does not perform as well as uranium for breeding in this type of reactor. Further test with different reflectors are planned for both U-Pu and Th-U fuel cycles. (Author)

  3. Improving the calculated core stability by the core nuclear design optimization

    International Nuclear Information System (INIS)

    Partanen, P.

    1995-01-01

    Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

  4. Enhanced E3 transitions and mixed configurations for core excited isomers in 210At and 211At

    International Nuclear Information System (INIS)

    Dracoulis, G.D.; Steed, C.A.; Byrne, A.P.; Poletti, S.J.; Stuchbery, A.E.; Bark, R.A.

    1986-09-01

    The lifetime and branching ratio of the 19 + isomer in 210 At have been measured. Its enhanced E3 decay and g-factor, and those of the related 39/2 - isomer in 211 At are compared with the results of a semi-empirical shell model calculation which includes couplings to the 3 - octupole vibration, resulting in mixed configurations. Lifetimes were also obtained for the 15 - isomer in 210 At, and he 29/2 + isomer in 209 At

  5. LMR design concepts for transuranic management in low sodium void worth cores

    International Nuclear Information System (INIS)

    Hill, R.N.

    1991-01-01

    The fuel cycle processing techniques and hard neutron spectrum of the integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also allow a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. (author)

  6. Intelligent stochastic optimization routine for in-core fuel cycle design

    International Nuclear Information System (INIS)

    Parks, G.T.

    1988-01-01

    Any reactor fuel management strategy must specify the fuel design, batch sizes, loading configurations, and operational procedures for each cycle. To permit detailed design studies, the complex core characteristics must necessarily be computer modeled. Thus, the identification of an optimal fuel cycle design represents an optimization problem with a nonlinear objective function (OF), nonlinear safety constraints, many control variables, and no direct derivative information. Most available library routines cannot tackle such problems; this paper introduces an intelligent stochastic optimization routine that can. There has been considerable interest recently in the application of stochastic methods to difficult optimization problems, based on the statistical mechanics algorithms originally attributed to Metropolis. Previous work showed that, in optimizing the performance of a British advanced gas-cooled reactor fuel stringer, a rudimentary version of the Metropolis algorithm performed as efficiently as the only suitable routine in the Numerical Algorithms Group library. Since then the performance of the Metropolis algorithm has been considerably enhanced by the introduction of self-tuning capabilities by which the routine adjusts its control parameters and search pattern as it progresses. Both features can be viewed as examples of artificial intelligence, in which the routine uses the accumulation of data, or experience, to guide its future actions

  7. Hybrid design method for air-core solenoid with axial homogeneity

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Li; Lee, Sang Jin [Uiduk University, Gyeongju (Korea, Republic of); Choi, Suk Jin [Institute for Basic Science, Daejeon (Korea, Republic of)

    2016-03-15

    In this paper, a hybrid method is proposed to design an air-core superconducting solenoid system for 6 T axial uniform magnetic field using Niobium Titanium (NbTi) superconducting wire. In order to minimize the volume of conductor, the hybrid optimization method including a linear programming and a nonlinear programming was adopted. The feasible space of solenoid is divided by several grids and the magnetic field at target point is approximated by the sum of magnetic field generated by an ideal current loop at the center of each grid. Using the linear programming, a global optimal current distribution in the feasible space can be indicated by non-zero current grids. Furthermore the clusters of the non-zero current grids also give the information of probable solenoids in the feasible space, such as the number, the shape, and so on. Applying these probable solenoids as the initial model, the final practical configuration of solenoids with integer layers can be obtained by the nonlinear programming. The design result illustrates the efficiency and the flexibility of the hybrid method. And this method can also be used for the magnet design which is required the high homogeneity within several ppm (parts per million)

  8. Interface Design Implications for Recalling the Spatial Configuration of Virtual Auditory Environments

    Science.gov (United States)

    McMullen, Kyla A.

    study also found that the presence of visual reference frames significantly increased recall accuracy. Additionally, the incorporation of drastic attenuation significantly improved environment recall accuracy. Through investigating the aforementioned concerns, the present study made initial footsteps guiding the design of virtual auditory environments that support spatial configuration recall.

  9. Aerodynamic Shape Optimization Design of Wing-Body Configuration Using a Hybrid FFD-RBF Parameterization Approach

    Science.gov (United States)

    Liu, Yuefeng; Duan, Zhuoyi; Chen, Song

    2017-10-01

    Aerodynamic shape optimization design aiming at improving the efficiency of an aircraft has always been a challenging task, especially when the configuration is complex. In this paper, a hybrid FFD-RBF surface parameterization approach has been proposed for designing a civil transport wing-body configuration. This approach is simple and efficient, with the FFD technique used for parameterizing the wing shape and the RBF interpolation approach used for handling the wing body junction part updating. Furthermore, combined with Cuckoo Search algorithm and Kriging surrogate model with expected improvement adaptive sampling criterion, an aerodynamic shape optimization design system has been established. Finally, the aerodynamic shape optimization design on DLR F4 wing-body configuration has been carried out as a study case, and the result has shown that the approach proposed in this paper is of good effectiveness.

  10. Comparison of 2-D Magnetic Designs of Selected Coil Configurations for the Next European Dipole (NED)

    CERN Document Server

    Toral, F; Felice, H; Fessia, Paolo; Loveridge, P W; Regis, Federico; Rochford, J; Sanz, S; Schwerg, Nikolai; Védrine, P; Völlinger, Christine

    2007-01-01

    The Next European Dipole (NED) activity is developing a high-performance Nb$_{3}$Sn wire (aiming at a non-copper critical current density of 1500 A/mm2 at 4.2 K and 15 T), within the framework of the Coordinated Accelerator Research in Europe (CARE) project. This activity is expected to lead to the fabrication of a large aperture, high field dipole magnet. In preparation for this phase, a Working Group on Magnet Design and Optimization (MDO) has been established to propose an optimal design. Other parallel Work Packages are concentrating on relevant topics, such as quench propagation simulation, innovative insulation techniques, and heat transfer measurements. In a first stage, the MDO Working Group has selected a number of coil configurations to be studied, together with salient parameters and features to be considered during the evaluation: the field quality, the superconductor efficiency, the conductor peak field, the stored magnetic energy, the Lorentz Forces and the fabrication difficulties. 2-D magnetic...

  11. Contribution to the build-up of a core calculation frame: comparison between ''diffusion'' and ''SPn'' operators on various configurations of the first N4 core

    International Nuclear Information System (INIS)

    Magat, Ph.

    1997-04-01

    The aim of this study is to compare two calculation methods implemented in the neutronic code CRONOS 2: the diffusion approximation and the SP n method. The APOLLO 2 code is used to build the multiparameter cross section libraries.The comparison is based on the first core of N4 type Chooz reactor. The rod worth and the power map have been calculated. Some recommendations about the SP n development order of flux are made and the results show that the diffusion calculations over-estimate the black rod efficiency up to 10%. (A.C.)

  12. Design of the core support and restraint structures for FFTF and CRBRP

    International Nuclear Information System (INIS)

    Sutton, H.G.; Rylatt, J.A.

    1977-12-01

    This paper presents and compares the design and fabrication of the FFTF and CRBRP reactor structures which support and restrain the reactor core assemblies. The fabrication of the core support structure (CSS) for the FFTF reactor was completed October 1972 and this paper discusses how the fabrication problems encountered with the FFTF were avoided in the subsequent design of the CRBR CSS. The radial core restraint structure of the FFTF was designed and fabricated such that an active system could replace the present passive system which is segmented and relies on the CSS core barrel for total structure integrity to maintain core geometry. The CRBR core restraint structure is designed for passive restraint only, and this paper discusses how the combined strengths of the restraint structure former rings and the CSS core barrel are utilized to maintain core geometry. Whereas the CSS for the FFTF interfaces directly with the reactor core assemblies, the CRBR CSS does not. A comparison is made on how intermediate structures in CRBR (inlet modules) provide the necessary design interfaces for supporting and providing flow distribution to the reactor core assemblies. A discussion is given on how the CRBR CSS satisfied the design requirements of the Equipment Specification, including thermal transient, dynamic and seismic loadings, and results of flow distribution testing that supported the CRBR design effort. The approach taken to simplify fabrication of the CRBR components, and a novel 20 inch deep narrow gap weld joint in the CSS are described

  13. Evaluation of applicability of 2D iron core model for two-limb configuration of GOLEM tokamak

    Czech Academy of Sciences Publication Activity Database

    Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Vondrášek, G.

    2013-01-01

    Roč. 88, 6-8 (2013), s. 835-838 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Tokamak * Ferromagnetic core * Integral method * Tokamak GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613002573#

  14. Earth Observatory Satellite system definition study. Report no. 3: Design/cost tradeoff studies. Appendix D: EOS configuration design data. Part 2: Data management system configuration

    Science.gov (United States)

    1974-01-01

    The Earth Observatory Satellite (EOS) data management system (DMS) is discussed. The DMS is composed of several subsystems or system elements which have basic purposes and are connected together so that the DMS can support the EOS program by providing the following: (1) payload data acquisition and recording, (2) data processing and product generation, (3) spacecraft and processing management and control, and (4) data user services. The configuration and purposes of the primary or high-data rate system and the secondary or local user system are explained. Diagrams of the systems are provided to support the systems analysis.

  15. Design of the zero power reactor core of Instituto de Energia Atomica, SP, Brazil

    International Nuclear Information System (INIS)

    Ferreira, Antonio Carlos de Almeida

    1974-01-01

    The main characteristics of a graphite moderated core of a critical assembly to be installed in the zero power reactor of the Instituto de Energia Atomica have been defined. Several simple geometric configurations have been selected and criticality studies have been made. The necessary quantity of fissile uranium has been calculated. (author)

  16. Configuration Management

    International Nuclear Information System (INIS)

    Morcos, A.; Taylor, H. S.

    1989-01-01

    This paper will briefly discuss the reason for and content of configuration management both for new plants and, when adapted, for older plants. It will then address three types of activities a utility may undertake as part of a nuclear CAM program and with which Sargent and Leyden has been actively involved. The first activity is a methodology for preparing design-basis documentation. The second is the identification of essential data required to be kept by the utility in support of the operation of a nuclear plant. The third activity is a computerized classification system of plant components, allowing ready identification of plant functional and physical characteristics. Plant configuration documentation describes plant components, the ways they arranged to interact, and the ways they are enabled to interact. Configuration management, on the other hand, is more than the control of such documentation. It is a dynamic process for ensuring that a plant configuration meets all relevant requirements for safety and economy, even while the configuration changes and even while the requirements change. Configuration management for a nuclear plant is so complex that it must be implemented in phases and modules. It takes advantage of and integrates existing programs. Managing complexity and streamlining the change process become important additional objectives of configuration management. The example activities fulfill essential goals of an overall CAM program: definition of design baseline, definition of essential plant data, and classification of plant components

  17. Application of core structural design guidelines in conceptual fuel pin design

    International Nuclear Information System (INIS)

    Patel, M.R.; Stephen, J.D.

    1979-01-01

    The paper describes an application of the Draft RDT Standards F9-7, -8, and -9 to conceptual design of Fast Breeder Reactor (FBR) fuel pins. The Standards are being developed to provide guidelines for structural analysis and design of the FBR core components which have limited ductility at high fluences and are not addressed by the prevalent codes. The development is guided by a national working group sponsored by the Division of Reactor Researcch and Technology of the Department of Energy. The development program summarized in the paper includes establishment of design margins consistent with the test data and component performance requirements, and application of the design rules in various design activities. The application program insures that the quantities required for proper application of the design rules are available from the analysis methods and test data, and that the use of the same design rules in different analysis tools used at different stages of a component design producees consistent results. This is illustrated in the paper by application of the design rules in the analysis methods developed for conceptual and more detailed designs of an FBR fuel pin

  18. Application of neural network to multi-dimensional design window search in reactor core design

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki

    1999-01-01

    In the reactor core design, many parametric survey calculations should be carried out to decide an optimal set of basic design parameter values. They consume a large amount of computation time and labor in the conventional way. To support design work, we investigate a procedure to search efficiently a design window, which is defined as feasible design parameter ranges satisfying design criteria and requirements, in a multi-dimensional space composed of several basic design parameters. The present method is applied to the neutronics and thermal hydraulics fields. The principle of the present method is to construct the multilayer neural network to simulate quickly a response of an analysis code through a training process, and to reduce computation time using the neural network without parametric study using analysis codes. To verify the applicability of the present method to the neutronics and the thermal hydraulics design, we have applied it to high conversion water reactors and examined effects of the structure of the neural network and the number of teaching patterns on the accuracy of the design window estimated by the neural network. From the results of the applications, a guideline to apply the present method is proposed and the present method can predict an appropriate design window in a reasonable computation time by following the guideline. (author)

  19. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  20. Pulsed air-core deflector-magnet design parameters

    International Nuclear Information System (INIS)

    Jason, A.J.; Cooper, R.K.; Liebman, A.D.; Blind, B.; Koelle, A.R.

    1983-01-01

    The response of air-core magnets to pulsed excitation is dependent on the pulse frequency spectrum because of fields produced by induced currents in the magnet structure. We discuss this phenomenon quantitatively in terms of magnet performance optimization

  1. Effect of rotor configuration on guyed tower and foundation designs and estimated costs for intermediate site horizontal axis wind turbines

    Science.gov (United States)

    Frederick, G. R.; Winemiller, J. R.; Savino, J. M.

    1982-01-01

    Three designs of a guyed cylindrical tower and its foundation for an intermediate size horizontal axis wind turbine generator are discussed. The primary difference in the three designs is the configuration of the rotor. Two configurations are two-blade rotors with teetering hubs - one with full span pitchable blades, the other with fixed pitch blades. The third configuration is a three-bladed rotor with a rigid hub and fixed pitch blades. In all configurations the diameter of the rotor is 38 meters and the axis of rotation is 30.4 meters above grade, and the power output is 200 kW and 400 kW. For each configuration the design is based upon for the most severe loading condition either operating wind or hurricane conditions. The diameter of the tower is selected to be 1.5 meters (since it was determined that this would provide sufficient space for access ladders within the tower) with guy rods attached at 10.7 meters above grade. Completing a design requires selecting the required thicknesses of the various cylindrical segments, the number and diameter of the guy rods, the number and size of soil anchors, and the size of the central foundation. The lower natural frequencies of vibration are determined for each design to ensure that operation near resonance does not occur. Finally, a cost estimate is prepared for each design. A preliminary design and cost estimate of a cantilever tower (cylindrical and not guyed) and its foundation is also presented for each of the three configurations.

  2. Thermal-hydraulic study of fixed bed nuclear reactor (FBNR), in FCC, BCC and pseudo-random configurations of the core through CFD method

    International Nuclear Information System (INIS)

    Luna, M.; Chavez, I.; Cajas, D.; Santos, R.

    2015-01-01

    The study of thermal-hydraulic performance of a fixed bed nuclear reactor (FBNR) core and the effect of the porosity was studied by the CFD method with 'SolidWorks' software. The representative sections of three different packed beds arrangements were analyzed: face-centered cubic (FCC), body-centered cubic (BCC), and a pseudo-random, with values of porosity of 0.28, 0.33 and 0.53 respectively. The minimum coolant flow required to avoid the phase change for each one of the configurations was determined. The results show that the heat transfer rate increases when the porosity value decreases, and consequently the minimum coolant flow in each configuration. The results of minimum coolant flow were: 728.51 kg/s for the FCC structure, 372.72 kg/s for the BCC, and 304.96 kg/s for the pseudo-random. Meanwhile, the heat transfer coefficients in each packed bed were 6480 W/m 2 *K, 3718 W/m 2 *K and 3042 W/m 2 *K respectively. Finally the pressure drop was calculated, and the results were 0.588 MPa for FCC configuration, 0.033 MPa for BCC and 0.017 MPa for the pseudo-random one. This means that with a higher porosity, the fluid can circulate easier because there are fewer obstacles to cross, so there are fewer energy losses. (authors)

  3. Uncertainty reevaluation of T/H parameters of HANARO core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hark Rho; Park, Cheol; Kim, Heo Nil; Chae, Hee Taek

    1999-03-01

    HANARO core was designed by statistical thermal design method which was generally applied to power plant design. However, reevaluation of core thermal margin reflecting design changes as well as experiences through commissioning and operation is necessary for safe operation of reactor. For this objective, the revision of data for T/H design parameters and the reevaluation of their uncertainties were performed. (Author). 30 refs., 7 figs.

  4. Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

    International Nuclear Information System (INIS)

    Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

    2000-01-01

    This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

  5. Adaptation and focusing of optode configurations for fluorescence optical tomography by experimental design methods.

    Science.gov (United States)

    Freiberger, Manuel; Clason, Christian; Scharfetter, Hermann

    2010-01-01

    Fluorescence tomography excites a fluorophore inside a sample by light sources on the surface. From boundary measurements of the fluorescent light, the distribution of the fluorophore is reconstructed. The optode placement determines the quality of the reconstructions in terms of, e.g., resolution and contrast-to-noise ratio. We address the adaptation of the measurement setup. The redundancy of the measurements is chosen as a quality criterion for the optodes and is computed from the Jacobian of the mathematical formulation of light propagation. The algorithm finds a subset with minimum redundancy in the measurements from a feasible pool of optodes. This allows biasing the design in order to favor reconstruction results inside a given region. Two different variations of the algorithm, based on geometric and arithmetic averaging, are compared. Both deliver similar optode configurations. The arithmetic averaging is slightly more stable, whereas the geometric averaging approach shows a better conditioning of the sensitivity matrix and mathematically corresponds more closely with entropy optimization. Adapted illumination and detector patterns are presented for an initial set of 96 optodes placed on a cylinder with focusing on different regions. Examples for the attenuation of fluorophore signals from regions outside the focus are given.

  6. Verification of the CENTRM Module for Adaptation of the SCALE Code to NGNP Prismatic and PBR Core Designs

    International Nuclear Information System (INIS)

    2014-01-01

    The generation of multigroup cross sections lies at the heart of the very high temperature reactor (VHTR) core design, whether the prismatic (block) or pebble-bed type. The design process, generally performed in three steps, is quite involved and its execution is crucial to proper reactor physics analyses. The primary purpose of this project is to develop the CENTRM cross-section processing module of the SCALE code package for application to prismatic or pebble-bed core designs. The team will include a detailed outline of the entire processing procedure for application of CENTRM in a final report complete with demonstration. In addition, they will conduct a thorough verification of the CENTRM code, which has yet to be performed. The tasks for this project are to: Thoroughly test the panel algorithm for neutron slowing down; Develop the panel algorithm for multi-materials; Establish a multigroup convergence 1D transport acceleration algorithm in the panel formalism; Verify CENTRM in 1D plane geometry; Create and test the corresponding transport/panel algorithm in spherical and cylindrical geometries; and, Apply the verified CENTRM code to current VHTR core design configurations for an infinite lattice, including assessing effectiveness of Dancoff corrections to simulate TRISO particle heterogeneity.

  7. Verification of the CENTRM Module for Adaptation of the SCALE Code to NGNP Prismatic and PBR Core Designs

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, Barry; Maldonado, Ivan

    2014-01-23

    The generation of multigroup cross sections lies at the heart of the very high temperature reactor (VHTR) core design, whether the prismatic (block) or pebble-bed type. The design process, generally performed in three steps, is quite involved and its execution is crucial to proper reactor physics analyses. The primary purpose of this project is to develop the CENTRM cross-section processing module of the SCALE code package for application to prismatic or pebble-bed core designs. The team will include a detailed outline of the entire processing procedure for application of CENTRM in a final report complete with demonstration. In addition, they will conduct a thorough verification of the CENTRM code, which has yet to be performed. The tasks for this project are to: Thoroughly test the panel algorithm for neutron slowing down; Develop the panel algorithm for multi-materials; Establish a multigroup convergence 1D transport acceleration algorithm in the panel formalism; Verify CENTRM in 1D plane geometry; Create and test the corresponding transport/panel algorithm in spherical and cylindrical geometries; and, Apply the verified CENTRM code to current VHTR core design configurations for an infinite lattice, including assessing effectiveness of Dancoff corrections to simulate TRISO particle heterogeneity.

  8. Verification of the CENTRM Module for Adaptation of the SCALE Code to NGNP Prismatic and PBR Core Designs

    International Nuclear Information System (INIS)

    Ganapol, Barry; Maldonodo, Ivan

    2014-01-01

    The generation of multigroup cross sections lies at the heart of the very high temperature reactor (VHTR) core design, whether the prismatic (block) or pebble-bed type. The design process, generally performed in three steps, is quite involved and its execution is crucial to proper reactor physics analyses. The primary purpose of this project is to develop the CENTRM cross-section processing module of the SCALE code package for application to prismatic or pebble-bed core designs. The team will include a detailed outline of the entire processing procedure for application of CENTRM in a final report complete with demonstration. In addition, they will conduct a thorough verification of the CENTRM code, which has yet to be performed. The tasks for this project are to: Thoroughly test the panel algorithm for neutron slowing down; Develop the panel algorithm for multi-materials; Establish a multigroup convergence 1D transport acceleration algorithm in the panel formalism; Verify CENTRM in 1D plane geometry; Create and test the corresponding transport/panel algorithm in spherical and cylindrical geometries; and, Apply the verified CENTRM code to current VHTR core design configurations for an infinite lattice, including assessing effectiveness of Dancoff corrections to simulate TRISO particle heterogeneity

  9. Design and analysis of three-layer-core optical fiber

    Science.gov (United States)

    Zheng, Siwen; Liu, Yazhuo; Chang, Guangjian

    2018-03-01

    A three-layer-core single-mode large-mode-area fiber is investigated. The three-layer structure in the core, which is composed of a core-index layer, a cladding-index layer, and a depression-index layer, could achieve a large effective area Aeff while maintaining an ultralow bending loss without deteriorating cutoff behaviors. The single-mode large mode area of 100 to 330 μm2 could be achieved in the fiber. The effective area Aeff can be further enlarged by adjusting the layer parameters. Furthermore, the bending property could be improved in this three-layer-core structure. The bending loss could decrease by 2 to 4 orders of magnitude compared with the conventional step-index fiber with the same Aeff. These characteristics of three-layer-core fiber suggest that it can be used in large-mode-area wide-bandwidth high-capacity transmission or high-power optical fiber laser and amplifier in optical communications, which could be used for the basic physical layer structure of big data storage, reading, calculation, and transmission applications.

  10. Multi-dimensional design window search system using neural networks in reactor core design

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki

    2000-02-01

    In the reactor core design, many parametric survey calculations should be carried out to decide an optimal set of basic design parameter values. They consume a large amount of computation time and labor in the conventional way. To support directly design work, we investigate a procedure to search efficiently a design window, which is defined as feasible design parameter ranges satisfying design criteria and requirements, in a multi-dimensional space composed of several basic design parameters. We apply the present method to the neutronics and thermal hydraulics fields and develop the multi-dimensional design window search system using it. The principle of the present method is to construct the multilayer neural network to simulate quickly a response of an analysis code through a training process, and to reduce computation time using the neural network without parametric study using analysis codes. The system works on an engineering workstation (EWS) with efficient man-machine interface for pre- and post-processing. This report describes the principle of the present method, the structure of the system, the guidance of the usages of the system, the guideline for the efficient training of neural networks, the instructions of the input data for analysis calculation and so on. (author)

  11. A multi-crucible core-catcher concept: Design considerations and basic results

    International Nuclear Information System (INIS)

    Szabo, I.

    1995-01-01

    A multi-crucible core-catcher concept to be implemented in new light water reactor containments has recently been proposed. This paper deals with conceptual design considerations and the various ways this type of core-catcher could be designed to meet requirements for reactor application. A systematic functional analysis of the multi-crucible core-catcher concept and the results of the preliminary design calculation are presented. Finally, the adequacy of the multi-crucible core-catcher concept for reactor application is discussed. (orig.)

  12. SCC500: next-generation infrared imaging camera core products with highly flexible architecture for unique camera designs

    Science.gov (United States)

    Rumbaugh, Roy N.; Grealish, Kevin; Kacir, Tom; Arsenault, Barry; Murphy, Robert H.; Miller, Scott

    2003-09-01

    A new 4th generation MicroIR architecture is introduced as the latest in the highly successful Standard Camera Core (SCC) series by BAE SYSTEMS to offer an infrared imaging engine with greatly reduced size, weight, power, and cost. The advanced SCC500 architecture provides great flexibility in configuration to include multiple resolutions, an industry standard Real Time Operating System (RTOS) for customer specific software application plug-ins, and a highly modular construction for unique physical and interface options. These microbolometer based camera cores offer outstanding and reliable performance over an extended operating temperature range to meet the demanding requirements of real-world environments. A highly integrated lens and shutter is included in the new SCC500 product enabling easy, drop-in camera designs for quick time-to-market product introductions.

  13. Status of reactor core design code system in COSINE code package

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Liu, Z.

    2014-01-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  14. Status of reactor core design code system in COSINE code package

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

    2014-07-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  15. Current directions in core-shell nanoparticle design

    Science.gov (United States)

    Schärtl, Wolfgang

    2010-06-01

    Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems.Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems

  16. Overview of the design of core restraint systems

    International Nuclear Information System (INIS)

    Heinecke, J.

    1984-01-01

    The optimization of the core restraint system is an important condition for the safe and reliable operation of a fast breeder reactor. For KNK II which is under successful operation and SNR 300 all requirements for safety and operation have been met with help of a ring type system. For SNR 2 the decision between the ring type system and the free standing core has to be done in the near future. Within these considerations the advantages of a ring type restraint system of limiting deflections during operation and limiting of possible movements under seismic conditions have to be balanced against the somewhat more complicated structure of the ring type restraint system

  17. Design features affecting dynamic behaviour of fast reactor cores

    International Nuclear Information System (INIS)

    Kayser, G.; Gouriou, A.

    1981-06-01

    The study of dynamic response of an LMFBR to normal and accidental transients needs first of all a simulation code taking into account all the important effects. The DYN-1 code aims at this target. It represents with a sufficiently accurate meshing the core in a 20 geometry for the thermal and reactivity effects, while the kinetics of this core are calculated with a point model. The primary pool, secondary loops, steam generator are also represented, as well as the control and protective systems. We give a short description of this code. Simpler codes are sometimes good enough for parametric studies

  18. CAREM 25: actual status of the core neutronic design. Calculation line

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    This work follows the one titled 'Criteria for the CAREM 25 reactor core design. Neutronic aspects' presented at this congress, gives in detail the typical values regarding the core defined at this point. Besides, the neutronic calculation line used for the CAREM 25 reactor design is presented. (Author) [es

  19. 75 FR 80571 - Core Principles and Other Requirements for Designated Contract Markets

    Science.gov (United States)

    2010-12-22

    ... Part II Commodity Futures Trading Commission 17 CFR Parts 1, 16, and 38 Core Principles and Other... CFR Parts 1, 16, and 38 RIN 3038-AD09 Core Principles and Other Requirements for Designated Contract... Principles 1. Subpart B--Designation as Contract Market 2. Subpart C--Compliance With Rules i. Proposed Sec...

  20. Analysis of design parameters for crosstalk cancellation filters applied to different loudspeaker configurations

    DEFF Research Database (Denmark)

    Lacouture Parodi, Yesenia; Rubak, Per

    2011-01-01

    for crosstalk cancellation filters applied to different loudspeaker configurations has not yet been addressed systematically. A study of three different inversion techniques applied to several loudspeaker arrangements is documented. Least-squares approximations in the frequency and time domains are evaluated...... along with a crosstalk canceler based on minimum-phase approximation with a frequency-independent delay. The three methods were applied to loudspeaker configurations with two channels and the least-squares approaches to configurations with four channels. Several different span angles and elevations were...

  1. Thermal hydraulic design of a hydride-fueled inverted PWR core

    International Nuclear Information System (INIS)

    Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

    2009-01-01

    An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

  2. Teaching to the Common Core by Design, Not Accident

    Science.gov (United States)

    Phillips, Vicki; Wong, Carina

    2012-01-01

    The Bill & Melinda Gates Foundation has created tools and supports intended to help teachers adapt to the Common Core State Standards in English language arts and mathematics. The tools seek to find the right balance between encouraging teachers' creativity and giving them enough guidance to ensure quality. They are the product of two years of…

  3. Common Core Standards and their Impact on Standardized Test Design

    NARCIS (Netherlands)

    Polleck, J.N.; Jeffery, J.V.

    2017-01-01

    With adoption of the Common Core (CCSS) in a majority of U.S. states came developmentof new high-stakes exams. Though researchers have investigated CCSS andrelated policies, less attention has been directed toward understanding how standardsare translated into testing. Due to the influence that

  4. Design studies for the Mark-III core of experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

    1979-08-01

    The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

  5. A Survey of Atom Interferometer Beam-Combination Configurations and Beam Splitter Designs

    National Research Council Canada - National Science Library

    Zhang, Xiaolei

    2005-01-01

    This report summarizes the state of the art of atom-interferometry experiments, with an emphasis on the beam-splitting and beam-combination configurations, as well as on the different choices of beam...

  6. Reference core design Mark-III of the experimental multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Watanabe, Takashi; Ishiguro, Okikazu; Kuroki, Syuzi

    1977-10-01

    The reactivity control system is one of the important items in reactor design, but it is much restricted by structural design of fuel element and pressure vessel in the experimental multi-purpose, high-temperature reactor. Preceding the first conceptual design of the reactor, therefore, the reactivity control system composed of control rod, burnable poison and reserve shutdown system in Mark-II design was re-studied, and several improvements were indicated. (1) The diameter of control rods must be as large as possible because it is impossible to increase the number of control rods. (2) The accuracy in estimation of the reactivity to be compensated with control rods is important because of the mutual interference of pair control rods with the twin configuration in a fuel element. (3) The improvement of core performance in burnup is accompanied by the reduction of design margin for control rods. (4) Increase of the reactivity to be compensated with the burnable poison leads to increase of the core reactivity recovery with burnup, and the assertion of the decrease for recovery of reactivity leads to increase of the temperature dependency of reactivity compensated with control rods. (5) Reduction of reactivity to be compensated with control rods is thus limited by cancellation of the effects in the reactivity recovery and the reactivity temperature dependency. (6) The reserve shutdown system can be designed with margin under the condition of excluding the reactivity of burnup from that to be compensated. (auth.)

  7. Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

    2012-06-01

    The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

  8. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    International Nuclear Information System (INIS)

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-01-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  9. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chisom Shawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  10. Core-Shell Al-Polytetrafluoroethylene (PTFE) Configurations to Enhance Reaction Kinetics and Energy Performance for Nanoenergetic Materials.

    Science.gov (United States)

    Wang, Jun; Qiao, Zhiqiang; Yang, Yuntao; Shen, Jinpeng; Long, Zhang; Li, Zhaoqian; Cui, Xudong; Yang, Guangcheng

    2016-01-04

    The energy performance of solid energetic materials (Al, Mg, etc.) is typically restricted by a natural passivation layer and the diffusion-limited kinetics between the oxidizer and the metal. In this work, we use polytetrafluoroethylene (PTFE) as the fluorine carrier and the shielding layer to construct a new type of nano-Al based fuels. The PTFE shell not only prevents nano-Al layers from oxidation, but also assists in enhancing the reaction kinetics, greatly improving the stability and reactivity of fuels. An in situ chemical vapor deposition combined with the electrical explosion of wires (EEW) method is used to fabricate core-shell nanostructures. Studies show that by controlling the stoichiometric ratio of the precursors, the morphology of the PTFE shell and the energy performance can be easily tuned. The resultant composites exhibit superior energy output characters than that of their physically mixed Al/PTFE counterparts. This synthetic strategy might provide a general approach to prepare other high-energy fuels (Mg, Si). © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. An integrated software system for core design and safety analyses: Cascade-3D

    International Nuclear Information System (INIS)

    Wan De Velde, A.; Finnemann, H.; Hahn, T.; Merk, S.

    1999-01-01

    The new Siemens program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of the most advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management. (authors)

  12. Design of low-loss and highly birefringent hollow-core photonic crystal fiber

    DEFF Research Database (Denmark)

    Roberts, Peter John; Williams, D.P.; Sabert, H.

    2006-01-01

    A practical hollow-core photonic crystal fiber design suitable for attaining low-loss propagation is analyzed. The geometry involves a number of localized elliptical features positioned on the glass ring that surrounds the air core and separates the core and cladding regions. The size of each...... feature is tuned so that the composite core-surround geometry is antiresonant within the cladding band gap, thus minimizing the guided mode field intensity both within the fiber material and at material / air interfaces. A birefringent design, which involves a 2-fold symmetric arrangement of the features...

  13. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  14. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  15. Design of the air-core transformer in spherical tokamak

    International Nuclear Information System (INIS)

    Wang Zhongtian; Jian Guangde; Li Fangzhu; Mao Guoping

    2002-01-01

    An ideal current distribution in the air-core transformer coils is obtained using variation principle. Climbing mountain method is utilized for optimizing the dimension and position of the real coils. Not only can the requirement of minimizing the stray field in the plasma region be guaranteed, but also integer turns for the coil can be realized. The latter brings a significant convenience to engineering

  16. Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

    International Nuclear Information System (INIS)

    Dozaki, Koji

    2007-01-01

    Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

  17. Advanced Intermediate Heat Transport Loop Design Configurations for Hydrogen Production Using High Temperature Nuclear Reactors

    International Nuclear Information System (INIS)

    Chang Oh; Cliff Davis; Rober Barner; Paul Pickard

    2005-01-01

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic evaluations and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various

  18. Computers in plasma physics: remote data access and magnetic configuration design

    International Nuclear Information System (INIS)

    Blackwell, B.D.; McMillan, B.F.; Searle, A.C.; Gardner, H.J.; Price, D.M.; Fredian, T.W.

    2000-01-01

    Full text: Two graphically intensive examples of the application of computers in plasma physics are described remote data access for plasma confinement experiments, and a code for real-time magnetic field tracing and optimisation. The application for both of these is the H-1NF National Plasma Fusion Research Facility, a Commonwealth Major National Research Facility within the Research School of Physical Science, Institute of Advanced Studies, ANU. It is based on the 'flexible' heliac stellarator H-1, a plasma confinement device in which the confining fields are generated solely by external conductors. These complex, fully three dimensional magnetic fields are used as examples for the magnetic design application, and data from plasma physics experiments are used to illustrate the remote access techniques. As plasma fusion experiments grow in size, increased remote access allows physicists to participate in experiments and data analysis from their home base. Three types of access will be described and demonstrated - a simple Java-based web interface, an example TCP client-server built around the widely used MDSPlus data system and the visualisation package IDL (RSI Inc), and a virtual desktop Environment (VNC: AT and T Research) that simulates terminals local to the plasma facility. A client server TCP/IP - web interface to the programmable logic controller that provides user interface to the programmable high power magnet power supplies is described. A very general configuration file allows great flexibility, and allows new displays and interfaces to be created (usually) without changes to the underlying C++ and Java code. The magnetic field code BLINE provides accurate calculation of complex magnetic fields, and 3D visualisation in real time, using a low cost multiprocessor computer and an OpenGL-compatible graphics accelerator. A fast, flexible multi-mesh interpolation method is used for tracing vacuum magnetic field lines created by arbitrary filamentary

  19. Feasibility of combining linear theory and impact theory methods for the analysis and design of high speed configurations

    Science.gov (United States)

    Brooke, D.; Vondrasek, D. V.

    1978-01-01

    The aerodynamic influence coefficients calculated using an existing linear theory program were used to modify the pressures calculated using impact theory. Application of the combined approach to several wing-alone configurations shows that the combined approach gives improved predictions of the local pressure and loadings over either linear theory alone or impact theory alone. The approach not only removes most of the short-comings of the individual methods, as applied in the Mach 4 to 8 range, but also provides the basis for an inverse design procedure applicable to high speed configurations.

  20. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

    2008-03-01

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  1. Mechanical design of core components for a high performance light water reactor with a three pass core

    International Nuclear Information System (INIS)

    Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

    2007-01-01

    Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

  2. Design factors affecting dynamic behaviour of fast reactor cores. UK review paper

    Energy Technology Data Exchange (ETDEWEB)

    Brindley, K W [National Nuclear Corporation Ltd., Risley, Warrington (United Kingdom); Perks, M A [United Kingdom Atomic Energy Authority, Risley, Warrington (United Kingdom)

    1982-01-01

    This paper summarises the consideration that has been given in the UK to the following factors that affect the dynamic behaviour of fast reactor cores: fuel design - Pu/u homogeneity, fuel expansion, fuel-clad gaps, uranium fraction. Structural response - CR supports, diagrid, sub-assembly bowing sodium expansion coefficients - low void cores including heterogenous cores. Calculational methods and models are outlined and some experimental results are discussed. (author)

  3. Polymer Design and Processing for Liquid-Core waveguides

    DEFF Research Database (Denmark)

    Sagar, Kaushal Shashikant

    precursor material. Upon attaining thermodynamically stable gyroid phase segregation, nanoporosity is induced by chemically removing PDMS, the so-called sacrificial block. The isotropic nanoporosity in the polymer is utilized in fabricating a novel type of waveguides for opto-fluidic applications, which we...... are spontaneously filled with water by capillary suction, forming the core, while the unmodified hydrophobic regions remain dry, forming the clad. Two types of photo-modification reactions are presented in this thesis: photo-oxidation and thiol-ene photo-clicking. The hydrophilicity is firstly induced by surface...

  4. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  5. Design Basis of Core Components and their Realization in the frame of the EPR'sTM Core Component Development

    International Nuclear Information System (INIS)

    Schebitz, Florian; Mekmouche, Abdelhalim

    2008-01-01

    Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI TM RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR TM , the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR TM . (authors)

  6. Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

    2016-02-15

    Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

  7. A method for designing configurations of nested monitoring wells near landfills

    Science.gov (United States)

    Hudak, Paul F.

    A method was devised for designing configurations of monitoring wells, consisting of vertically nested intakes in boreholes. The network-design method involves analyzing a subset of potential contaminant plumes emerging from the downgradient margin of a landfill. Plume widths are evaluated along selected equipotential lines and compared to the lengths of those lines. The method was applied to a 32-ha solid-waste landfill in Tarrant County, Texas, USA. Sixtynine potential source nodes were considered. A 15-borehole network devised by the method registered 93 detections in total, detecting all 69 model-generated plumes by at least one borehole. Based on an enumeration procedure, a minimum of 10 boreholes was needed to detect all of the model-generated plumes. However, the less conservative 10-borehole network had little capability for backup detection. An existing monitoring network of seven downgradient wells detected only 38 model-generated plumes. Results of this study illustrate a practical need for structured approaches to designing detection-based groundwater-monitoring configurations. Résumé Une méthode a été développée pour fournir les caractéristiques de puits de surveillance, avec des points de prélèvements superposés en forage. La méthode de réalisation du réseau s'appuie sur l'analyse d'un ensemble de panaches de pollution potentiels provenant du bord en aval d'une décharge. Les largeurs de panache sont estimées le long d'isopièzes sélectionnées et sont comparées à leur longueur. Cette méthode a été appliquée à une décharge de déchets solides couvrant 32ha, dans le canton de Tarrant (Texas, Etats-Unis). 69 noeuds de source potentielle de pollution ont été pris en compte. Un réseau de 15 forages, défini par la méthode, a enregistré au total 93 alarmes, détectant les 69 panaches simulés dans au moins un forage. Une procédure de dénombrement précise qu'un minimum de 10 forages est nécessaire pour détecter tous les

  8. Effect of the design change of the LSSBP on core flow distribution of APR+ Reactor

    International Nuclear Information System (INIS)

    Kim, Kihwan; Euh, Dong-Jin; Choi, Hae-Seob; Kwon, Tae-Soon

    2014-01-01

    The uniform core inlet flow distribution of an Advanced Power Reactor Plus (APR+) is required to prevent the failure rate of the HIPER fuel assembly and improve the core thermal margin. KEPCO-E and C and KAERI proposed a design change of the Lower Support Structure Bottom Plate (LSSBP), since the core flow rates were intense near the outer region of the intact LSSBP in a previous study. In this study, an experiment was carried out to evaluate the effect of the design change of the LSSBP on the core flow distribution using the APR+ Core Flow and Pressure (ACOP) test facility. The results showed great improvement on the core flow distribution under a 4-pump balanced flow condition. Under the 4-pump balanced flow condition, fifteen tests were repeated using the ACOP test facility to verify the effect of the 50% blocked flow area at the outer region of the LSSBP on the core inlet flow distribution. The profiles of the core inlet mass flow rates were analyzed using ensemble averaged values, and compared with that of the intact LSSBP. The results showed great improvement for the overall core region. The change in design of the LSSBP is expected to improve the hydraulic performance of an APR+ reactor

  9. Design review report for rotary mode core sample truck (RMCST) modifications for flammable gas tanks, preliminary design

    International Nuclear Information System (INIS)

    Corbett, J.E.

    1996-02-01

    This report documents the completion of a preliminary design review for the Rotary Mode Core Sample Truck (RMCST) modifications for flammable gas tanks. The RMCST modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to validate basic design assumptions and concepts to support a path forward leading to a final design. The conclusion reached by the review committee was that the design was acceptable and efforts should continue toward a final design review

  10. Preliminary study on flexible core design of super FBR with multi-axial fuel shuffling

    International Nuclear Information System (INIS)

    Sukarman; Yamaji, Akifumi; Someya, Takayuki; Noda, Shogo

    2017-01-01

    Preliminary study has been conducted on developing a new flexible core design concept for the Supercritical water-cooled Fast Breeder Reactor (Super FBR) with multi-axial fuel shuffling. The proposed new concept focuses on the characteristic large axial coolant density change in supercritical water cooled reactors (SCWRs) when the coolant inlet temperature is below the pseudocritical point and large coolant enthalpy rise is taken in the core for achieving high thermal efficiency. The aim of the concept is to attain both the high breeding performance and good thermal-hydraulic performance at the same time. That is, short Compound System Doubling Time (CSDT) for high breeding, large coolant enthalpy rise for high thermal efficiency, and large core power. The proposed core concept consists of horizontal layers of mixed oxide (MOX) fuels and depleted uranium (DU) blanket layers at different elevation levels. Furthermore, the upper core and the lower core are separated and independent fuel shuffling schemes in these two core regions are considered. The number of fuel batches and fuel shuffling scheme of the upper core were changed to investigate influence of multi-axial fuel shuffling on the core characteristics. The core characteristics are evaluated with-three-dimensional diffusion calculations, which are fully-coupled with thermal-hydraulics calculations based on single channel analysis model. The results indicate that the proposed multi-axial fuel shuffling scheme does have a large influence on CSDT. Further investigations are necessary to develop the core concept. (author)

  11. Performance monitoring of different module technologies and design configurations of PV systems in South Africa

    CSIR Research Space (South Africa)

    Serameng, T

    2016-06-01

    Full Text Available Africa. The goal of this work was to analyse and report on the performance of PV systems by evaluating the energy output of various PV system configurations and module technologies in the South African (southern hemisphere) climatic conditions. To achieve...

  12. Design of Multi-core Fiber Patch Panel for Space Division Multiplexing Implementations

    DEFF Research Database (Denmark)

    Gonzalez, Luz E.; Morales, Alvaro; Rommel, Simon

    2018-01-01

    A multi-core fiber (MCF) patch panel was designed, allowing easy coupling of individual signals to and from a 7-core MCF. The device was characterized, measuring insertion loss and cross talk, finding highest insertion loss and lowest crosstalk at 1300 nm with values of 9.7 dB and -36.5 d...

  13. Designing Class Activities to Meet Specific Core Training Competencies: A Developmental Approach

    Science.gov (United States)

    Guth, Lorraine J.; McDonnell, Kelly A.

    2004-01-01

    This article presents a developmental model for designing and utilizing class activities to meet specific Association for Specialists in Group Work (ASGW) core training competencies for group workers. A review of the relevant literature about teaching group work and meeting core training standards is provided. The authors suggest a process by…

  14. Design of multi-core fiber patch panel for space division multiplexing implementations

    NARCIS (Netherlands)

    González, Luz E.; Morales, Alvaro; Rommel, Simon; Jørgensen, Bo F.; Porras-Montenegro, N.; Tafur Monroy, Idelfonso

    2018-01-01

    A multi-core fiber (MCF) patch panel was designed, allowing easy coupling of individual signals to and from a 7-core MCF. The device was characterized, measuring insertion loss and cross talk, finding highest insertion loss and lowest crosstalk at 1300 nm with values of 9.7 dB and -36.5 dB

  15. Thermal-hydraulic mixing in the split-core ANS reactor design

    International Nuclear Information System (INIS)

    Dorning, R.J.J.

    1988-01-01

    A design has been proposed for the advanced neutron source (ANS) reactor that incorporates a split core, one purpose of which is to create a mixing plenum between the upper and lower cores. It was hoped that in addition to introducing various desirable neutronics features, such as decreasing the fast neutron flux contamination of thermal and cold neutron beams located in the reactor midplane, this mixing plenum would make possible higher operating powers by lowering the maximum core temperature. This lower temperature was to be achieved as a result of the mixing, of the hot D 2 O coolant exiting the upper-core channels, and the cold D 2 O leaving the large upper core bypass. It was expected that this mixing would bring about a significantly reduced lower core maximum coolant inlet temperature. The authors have carried out large-scale computer calculations to determine the extent to which this mixing occurs in current split-core design geometry, which does not incorporate baffles, mixing devices, or other design features introduced to enhance mixing. The large-scale self-consistent calculations summarized here indicate that innovative design ideas to enhance mixing will be necessary if the split-core concept is to achieve the amount of thermal mixing needed to make possible significantly higher power operation and corresponding higher flux sources

  16. Melt spreading code assessment, modifications, and application to the EPR core catcher design

    International Nuclear Information System (INIS)

    Farmer, M.T.

    2009-01-01

    The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial validation consisted

  17. FFTF reload core nuclear design for increased experimental capability

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Nelson, J.V.; Dobbin, K.D.; Bennett, R.A.

    1976-01-01

    In anticipation of continued growth in the FTR experimental irradiations program, the enrichments for the next batches of reload driver fuel to be manufactured have been increased to provide a substantially enlarged experimental reactivity allowance. The enrichments for these fuel assemblies, termed ''Cores 3 and 4,'' were selected to meet the following objectives and constraints: (1) maintain a reactor power capability of 400 MW (based on an evaluation of driver fuel centerline melting probability at 15 percent overpower); (2) provide a peak neutron flux of nominally 7 x 10 15 n/cm 2 -sec, with a minimum acceptable value of 95 percent of this (i.e., 6.65 x 10 15 n/cm 2 -sec); and (3) provide the maximum experimental reactivity allowance that is consistent with the above constraints

  18. Proposal of C-core Type Transverse Flux Motor for Ship Propulsion – Increasing Torque Density by Dense Stator Configuration

    Directory of Open Access Journals (Sweden)

    Y. Yamamoto

    2014-02-01

    Full Text Available Electric ship propulsion system has been drawing attention as a solution for savings in energy and maintenance costs. The system is mainly composed of motor, converter and gearbox and required for high torque at low speed. In this situation, transverse flux motors (TFMs have been proposed to fulfill the low-speed high-torque characteristic due to suitable for short pole pitch and large number of poles to increase torque output. In this trend, we have proposed C-core type motors taking advantage of TFMs’ structure. In this manuscript, a simple design method based on the magnetic-circuit theory and simple modeling of the motor is proposed to search a design parameter for maximizing torque as a pre-process of numerical study. The method takes into consideration the effects of magnetic leakage flux, magnetic saturation and pole-core combination in accordance with the systematic theory. The simple modeling is conducted based on a dense armature structure in previous axial flux motors (AFMs applied to the new motor design. The validity of the method is verified by 3-D finite element analysis (FEA and relative error is at most 20%. The minimalist design is shown to be advantageous for effective use in 3-D FEA. As a detailed design by the FEA, high torque density and low cogging to output ratio can be achieved simultaneously in the proposed machine.

  19. Design of modular coils for a quasi-axisymmetric stellarator with a flexible control of the magnetic field configuration

    International Nuclear Information System (INIS)

    Shimizu, A.; Okamura, S.; Isobe, M.; Suzuki, C.; Nishimura, S.; Watari, T.; Matsuoka, K.

    2002-08-01

    A design of the modular coil system for CHS-qa has been made for the plasma configuration '2b32' with the aspect ratio 3.2. The magnetic field strength and the major radius are 1.5 T and 1.5 m, respectively. The normal component of magnetic field produced by the modular coils is minimized on the plasma boundary to obtain the optimum coil design. We put engineering constraint on the distance between adjacent modular coils and the radius of coil curvature. The dependence of the residual normal component of the field on these conditions is examined, and the realistic values for them are selected. Additional coils to control various properties of the magnetic field configuration (the rotational transform, the magnetic well depth, etc.) have been designed and a flexibility of the magnetic field configuration is realized. For the case that the rotational transform crosses the low-order rational value resulting in magnetic islands, the residues of islands are evaluated with which a further improvement of coil design can be made to eliminate magnetic islands. (author)

  20. A hybrid method for in-core optimization of pressurized water reactor reload core design

    International Nuclear Information System (INIS)

    Stevens, J.G.

    1995-05-01

    The objective of this research is the development of an accurate, practical, and robust method for optimization of the design of loading patterns for pressurized water reactors, a nonlinear, non-convex, integer optimization problem. The many logical constraints which may be applied during the design process are modeled herein by a network construction upon which performance objectives and safety constraints from reactor physics calculations are optimized. This thesis presents the synthesis of the strengths of previous algorithms developed for reload design optimization and extension of robustness through development of a hybrid liberated search algorithm. Development of three independent methods for reload design optimization is presented: random direct search for local improvement, liberated search by simulated annealing, and deterministic search for local improvement via successive linear assignment by branch and bound. Comparative application of the methods to a variety of problems is discussed, including an exhaustive enumeration benchmark created to allow comparison of search results to a known global optimum for a large scale problem. While direct search and determinism are shown to be capable of finding improvement, only the liberation of simulated annealing is found to perform robustly in the non-convex design spaces. The hybrid method SHAMAN is presented. The algorithm applies: determinism to shuffle an initial solution for satisfaction of heuristics and symmetry; liberated search through simulated annealing with a bounds cooling constraint treatment; and search bias through relational heuristics for the application of engineering judgment. The accuracy, practicality, and robustness of the SHAMAN algorithm is demonstrated through application to a variety of reload loading pattern optimization problems

  1. A safety design approach for sodium cooled fast reactor core toward commercialization in Japan

    International Nuclear Information System (INIS)

    Kubo, Shigenobu

    2012-01-01

    JAEA’s safety approach for SFR core design is based on defence‐in‐depth concept, which includes DBAs and DECs (prevention and mitigation): • The reactor core is designed to have inherent reactivity feedback characteristics with negative power coefficient. • Operation temperature range is set sufficiently below the coolant boiling temperature so as to avoid coolant boiling against anticipated operational occurrences and DBAs. • If the plant state deviates from operational states, the safe reactor shutdown is achieved by automatic insertion of control rods. 2 active reactor shutdown systems are provided. • Failure of active reactor shutdown is assumed in a design extension condition . Passive shutdown capability is provided by SASS under such condition. • As a design extension condition, core disruptive accident is assumed. In order to prevent severe mechanical energy release which might cause containment function failure, core sodium void worth is limited below 6 dollars and molten fuel discharge capability is utilized by FAIDUS. (author)

  2. Optimal Design and Analysis of the Stepped Core for Wireless Power Transfer Systems

    Directory of Open Access Journals (Sweden)

    Xiu Zhang

    2016-01-01

    Full Text Available The key of wireless power transfer technology rests on finding the most suitable means to improve the efficiency of the system. The wireless power transfer system applied in implantable medical devices can reduce the patients’ physical and economic burden because it will achieve charging in vitro. For a deep brain stimulator, in this paper, the transmitter coil is designed and optimized. According to the previous research results, the coils with ferrite core can improve the performance of the wireless power transfer system. Compared with the normal ferrite core, the stepped core can produce more uniform magnetic flux density. In this paper, the finite element method (FEM is used to analyze the system. The simulation results indicate that the core loss generated in the optimal stepped ferrite core can reduce about 10% compared with the normal ferrite core, and the efficiency of the wireless power transfer system can be increased significantly.

  3. Design of an Oximeter Based on LED-LED Configuration and FPGA Technology

    Directory of Open Access Journals (Sweden)

    Radovan Stojanovic

    2013-01-01

    Full Text Available A fully digital photoplethysmographic (PPG sensor and actuator has been developed. The sensing circuit uses one Light Emitting Diode (LED for emitting light into human tissue and one LED for detecting the reflectance light from human tissue. A Field Programmable Gate Array (FPGA is used to control the LEDs and determine the PPG and Blood Oxygen Saturation (SpO2. The configurations with two LEDs and four LEDs are developed for measuring PPG signal and Blood Oxygen Saturation (SpO2. N-LEDs configuration is proposed for multichannel SpO2 measurements. The approach resulted in better spectral sensitivity, increased and adjustable resolution, reduced noise, small size, low cost and low power consumption.

  4. Design of an oximeter based on LED-LED configuration and FPGA technology.

    Science.gov (United States)

    Stojanovic, Radovan; Karadaglic, Dejan

    2013-01-04

    A fully digital photoplethysmographic (PPG) sensor and actuator has been developed. The sensing circuit uses one Light Emitting Diode (LED) for emitting light into human tissue and one LED for detecting the reflectance light from human tissue. A Field Programmable Gate Array (FPGA) is used to control the LEDs and determine the PPG and Blood Oxygen Saturation (S(p)O(2)). The configurations with two LEDs and four LEDs are developed for measuring PPG signal and Blood Oxygen Saturation (S(p)O(2)). N-LEDs configuration is proposed for multichannel S(p)O(2) measurements. The approach resulted in better spectral sensitivity, increased and adjustable resolution, reduced noise, small size, low cost and low power consumption.

  5. Study in stationary state of the subcriticality of intermediate configurations of core in the reloading process of a BWR; Estudio en estado estacionario de la subcriticidad de configuraciones intermedias de nucleo en el proceso de recarga de un BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

    In this work is carried out the simulation in three dimensions with the COREMASTER-PRESTO code, of the behavior of the reactor core in different stages of the change process of fuel assemblies. To carry out the simulation, this code requires of a database of nuclear parameters that includes those that can associate to the areas of an assemblies that they don't contain fuel and in its place there is moderator. These nuclear parameters are calculated with the AURORA-HELIOS-ZENITH-TABGEN system. One of the approaches that were carried out consisted on designing a 'water assemble', that is to say, an axial arrangement of 25 'water cells'. To obtain the appropriate 'water cell' its were carried out some selective test cases, since it presents in two cases the necessity to find an enough minimum value of fissile material for the correct execution of HELIOS, firstly, and later on COREMASTER-PRESTO. In the first case, the situation is solved when placing symmetrically 6 bars with natural uranium in the lateral areas of the cell; with that which the value of k{sub inf} of 0.1592 is obtained in the calculations with the HELIOS code in the cold condition to zero power (CZP), and 0% of vacuums. For the second case the cell includes symmetrically 28 bars with natural uranium, and the k{sub inf} value is 0.45290. These values are the maximum through the life of the 'cell.' As part of the activities that are developed during the fuel substitution, this the one of evaluating the subcriticality of the core each determined number of substitution movements. The obtained results when evaluating the k-effective in cold condition, in 5 different intermediate core configurations, as the loading process of the fuel advances are presented. To make the evaluation with CM-PRESTO in each configuration it was proceeded to complete the rest of the 444 assemblies with the one denominated 'water assemble'. In all the evaluated cases the

  6. Legal Protection on IP Cores for System-on-Chip Designs

    Science.gov (United States)

    Kinoshita, Takahiko

    The current semiconductor industry has shifted from vertical integrated model to horizontal specialization model in term of integrated circuit manufacturing. In this circumstance, IP cores as solutions for System-on-Chip (SoC) have become increasingly important for semiconductor business. This paper examines to what extent IP cores of SoC effectively can be protected by current intellectual property system including integrated circuit layout design law, patent law, design law, copyright law and unfair competition prevention act.

  7. Design of a reactor core in the Oma Full MOX-ABWR

    International Nuclear Information System (INIS)

    Hama, Teruo

    1999-01-01

    The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

  8. A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

    International Nuclear Information System (INIS)

    Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

    2009-01-01

    The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

  9. Low-stress mounting configuration design for large aperture laser transport mirror

    Science.gov (United States)

    Zhang, Zheng; Quan, Xusong; Yao, Chao; Wang, Hui

    2016-10-01

    TM1-6S1 large aperture laser transport mirror is a crucial optical unit of high power solid-state laser in the Inertial Confinement Fusion (ICF) facility. This article focuses on the low-stress and precise mounting method of large-aperture mirror. Based on the engineering practice of SG-III, the state-of-the-art and key problems of current mounting configuration are clarified firstly. Subsequently, a brand new low-stress mounting configuration with flexure supports is proposed. Opto-mechanical model of the mirror under mounting force is built up with elastic mechanics theory. Further, numerical methods and field tests are employed to verify the favorable load uniform capacity and load adjust capacity of flexure supports. With FEM, the relation between the mounting force from new configuration and the mirror surface distortion (wavefront error) is clarified. The novel mounting method of large aperture optics could be not only used on this laser transport mirror, but also on the other transmission optics and large crystals in ICF facilities.

  10. Design and configuration of VME EPICS driver for He RFQ LLRF control system

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Tae-Sung; Jeong, Hae-Seong; Kim, Seong-Gu; Song, Young-Gi; Kim, Han-Sung; Seol, Kyung-Tae; Kwon, Hyeok-Jung; Cho, Yong-Sub [KOMAC, Gyeongju (Korea, Republic of)

    2015-05-15

    In Helium Radio-Frequency Quadrupole (He RFQ) development, the role of the high-power Radio-Frequency (RF) is very important because it is responsible for stable delivery and efficient acceleration of the beam. Since the amplitude control requirements of LLRF system are ±1 % (amplitude), we need a precise remote control system for this reason. This system is referred to as Low-Level RF (LLRF) control system. This paper describes the basic configuration tasks performed by hardware side and the software side to build the LLRF control system, and describes the future work of the He RFQ LLRF control system based on this paper. LLRF control system development at the He RFQ development stage is important. LLRF control system development requires the exact configuration of hardware and software. For each of the Layer configuration is completed on the software side and hardware modules: vxworks operating system installation, EPICS BASE compilation, module source code compiled, object file loading and execution on vxworks, EPICS IOC operation check, etc.

  11. Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

    Energy Technology Data Exchange (ETDEWEB)

    Benoît, Fabrice, E-mail: fabrice-2.benoit@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Allegretti, Ludovic [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Aumeunier, Marie-Hélène [OPTIS, ZE de La Farlède, F-83078 Toulon Cedex 9 (France); Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Robert, Julien [SOFYNE, F-69800 Saint Priest (France)

    2014-10-15

    Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule.

  12. Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

    International Nuclear Information System (INIS)

    Benoît, Fabrice; Allegretti, Ludovic; Aumeunier, Marie-Hélène; Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène; Robert, Julien

    2014-01-01

    Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule

  13. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  14. Core design options for high conversion BWRs operating in Th–233U fuel cycle

    International Nuclear Information System (INIS)

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2013-01-01

    Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

  15. LMR design concepts for transuranic management in low sodium void worth cores

    International Nuclear Information System (INIS)

    Hill, R.N.

    1991-01-01

    The fuel cycle processing techniques and hard neuron spectrum of the Integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also show a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. 13 refs., 1 fig., 4 tabs

  16. Introduction to Open Core Protocol Fastpath to System-on-Chip Design

    CERN Document Server

    Schwaderer, W David

    2012-01-01

    This book introduces Open Core Protocol (OCP), not as a conventional hardware communications protocol but as a meta-protocol: a means for describing and capturing the communications requirements of an IP core, and mapping them to a specific set of signals with known semantics.  Readers will learn the capabilities of OCP as a semiconductor hardware interface specification that allows different System-On-Chip (SoC) cores to communicate.  The OCP methodology presented enables intellectual property designers to design core interfaces in standard ways. This facilitates reusing OCP-compliant cores across multiple SoC designs which, in turn, drastically reduces design times, support costs, and overall cost for electronics/SoCs. Provides a comprehensive introduction to Open Core Protocol, which is more accessible than the full specification; Designed as a hands-on, how-to guide to semiconductor design; Includes numerous, real “usage examples” which are not available in the full specification; Integrates coverag...

  17. Specialists' meeting on design features affecting a dynamic behaviour of fast reactor cores. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-01-01

    The purpose of the meeting was to review and discuss the effects induced by changes in some design characteristics on overall performances and transient behaviour of fast reactor cores. The main topics discussed in the four technical sessions were: National Review Presentations. Identification of the key issues to be considered in the following sessions; Effects of design changes on performance characteristics. Kinetics models and codes; Evaluation and interpretation of reactivity coefficients. Kinetics calculations for restrained and free-standing cores; Comparison of the dynamic behaviour of homogeneous and heterogeneous cores.

  18. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  19. Specialists' meeting on design features affecting a dynamic behaviour of fast reactor cores. Summary report

    International Nuclear Information System (INIS)

    1982-01-01

    The purpose of the meeting was to review and discuss the effects induced by changes in some design characteristics on overall performances and transient behaviour of fast reactor cores. The main topics discussed in the four technical sessions were: National Review Presentations. Identification of the key issues to be considered in the following sessions; Effects of design changes on performance characteristics. Kinetics models and codes; Evaluation and interpretation of reactivity coefficients. Kinetics calculations for restrained and free-standing cores; Comparison of the dynamic behaviour of homogeneous and heterogeneous cores

  20. A design method to isothermalize the core of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takano, M.; Sawa, K.

    1987-01-01

    A practical design method is developed to isothermalize the core of block-type high-temperature gas-cooled reactors (HTGRs). Isothermalization plays an important role in increasing the design margin on fuel temperature. In this method, the fuel enrichment and the size and boron content of the burnable poison rod are determined over the core blockwise so that the axially exponential and radially flat power distribution are kept from the beginning to the end of core life. The method enables conventional HTGRs to raise the outlet gas temperature without increasing the maximum fuel temperature

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

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

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

  2. Status of the Astrid core at the end of the pre-conceptual design phase 1

    International Nuclear Information System (INIS)

    Chenaud, Ms.; Devictor, N.; Mignot, G.; Varaine, F.; Venard, C.; Martin, L.; Phelip, M.; Lorenzo, D.; Serre, F.; Bertrand, F.; Alpy, N.; Le-Flem, M.; Gavoille, P.; Lavastre, R.; Richard, P.; Verrier, D.; Schmitt, D.

    2013-01-01

    Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase. (authors)

  3. Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

    2016-01-15

    Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

  4. Reactor core design calculations and fuel management in PWR; Izracun projekta sredice in upravljanja z forivom tlacnovodnega reaktorja

    Energy Technology Data Exchange (ETDEWEB)

    Ravnik, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    Computer programs and methods developed at J. Stefan Institute for nuclear core design of Krsko NPP are treated. development, scope, verification and organisation of core design procedure are presented. The core design procedure is applicable to any NPP of PWR type. (author)

  5. Multiobjective optimization for design of multifunctional sandwich panel heat pipes with micro-architected truss cores

    International Nuclear Information System (INIS)

    Roper, Christopher S.

    2011-01-01

    A micro-architected multifunctional structure, a sandwich panel heat pipe with a micro-scale truss core and arterial wick, is modeled and optimized. To characterize multiple functionalities, objective equations are formulated for density, compressive modulus, compressive strength, and maximum heat flux. Multiobjective optimization is used to determine the Pareto-optimal design surfaces, which consist of hundreds of individually optimized designs. The Pareto-optimal surfaces for different working fluids (water, ethanol, and perfluoro(methylcyclohexane)) as well as different micro-scale truss core materials (metal, ceramic, and polymer) are determined and compared. Examination of the Pareto fronts allows comparison of the trade-offs between density, compressive stiffness, compressive strength, and maximum heat flux in the design of multifunctional sandwich panel heat pipes with micro-scale truss cores. Heat fluxes up to 3.0 MW/m 2 are predicted for silicon carbide truss core heat pipes with water as the working fluid.

  6. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  7. Uncertainty reduction requirements in cores designed for passive reactivity shutdown

    International Nuclear Information System (INIS)

    Wade, D.C.

    1988-01-01

    The first purpose of this paper is to describe the changed focus of neutronics accuracy requirements existing in the current US advanced LMR development program where passive shutdown is a major design goal. The second purpose is to provide the background and rationale which supports the selection of a formal data fitting methodology as the means for the application of critical experiment measurements to meet these accuracy needs. 6 refs., 1 fig., 2 tabs

  8. Development of core technology for KNGR system design

    International Nuclear Information System (INIS)

    Lee, K. Y; Park, J. K.; Ham, C. S.

    2002-05-01

    For evaluating the design issues in the advanced MMI, we have studied the selection of evaluation issues, evaluation schedules, and the application method of evaluation results from the beginning of APR1400 MMI design project. For producing input for human reliability analysis, we proposed a new error taxonomy and approach to explain the new human error potential. In order to evaluate whether or not the lifetime of APR1400 I and C system meet 60 years, the lifetime of the system has been calculated by using failure rates in this research. In order to qualify the software of APR1400 protection system, the detail scope and contents are to develop the methods, to evaluate the design results using the methods, and to establish a database for the evaluation. In external vessel cooling strategy to maintain the reactor vessel integrity in a severe accident of the APR 1400, an evaluation methodology was developed, and the failure possibility of penetrations and failure mode of the reactor vessel penetrations have been estimated through experiments and analysis. Two types of experiments have been performed using alumina melt as a simulant and a sustained heating method using an inductor, and their results have been evaluated along with APR 1400 using the LILAC, the FLUENT, and the ABAQUS computer codes

  9. Automated Integration of Dedicated Hardwired IP Cores in Heterogeneous MPSoCs Designed with ESPAM

    Directory of Open Access Journals (Sweden)

    Ed Deprettere

    2008-06-01

    Full Text Available This paper presents a methodology and techniques for automated integration of dedicated hardwired (HW IP cores into heterogeneous multiprocessor systems. We propose an IP core integration approach based on an HW module generation that consists of a wrapper around a predefined IP core. This approach has been implemented in a tool called ESPAM for automated multiprocessor system design, programming, and implementation. In order to keep high performance of the integrated IP cores, the structure of the IP core wrapper is devised in a way that adequately represents and efficiently implements the main characteristics of the formal model of computation, namely, Kahn process networks, we use as an underlying programming model in ESPAM. We present details about the structure of the HW module, the supported types of IP cores, and the minimum interfaces these IP cores have to provide in order to allow automated integration in heterogeneous multiprocessor systems generated by ESPAM. The ESPAM design flow, the multiprocessor platforms we consider, and the underlying programming (KPN model are introduced as well. Furthermore, we present the efficiency of our approach by applying our methodology and ESPAM tool to automatically generate, implement, and program heterogeneous multiprocessor systems that integrate dedicated IP cores and execute real-life applications.

  10. A comprehensive review of microbial electrolysis cells (MEC reactor designs and configurations for sustainable hydrogen gas production

    Directory of Open Access Journals (Sweden)

    Abudukeremu Kadier

    2016-03-01

    Full Text Available Hydrogen gas has tremendous potential as an environmentally acceptable energy carrier for vehicles. A cutting edge technology called a microbial electrolysis cell (MEC can achieve sustainable and clean hydrogen production from a wide range of renewable biomass and wastewaters. Enhancing the hydrogen production rate and lowering the energy input are the main challenges of MEC technology. MEC reactor design is one of the crucial factors which directly influence on hydrogen and current production rate in MECs. The rector design is also a key factor to up-scaling. Traditional MEC designs incorporated membranes, but it was recently shown that membrane-free designs can lead to both high hydrogen recoveries and production rates. Since then multiple studies have developed reactors that operate without membranes. This review provides a brief overview of recent advances in research on scalable MEC reactor design and configurations.

  11. Study on design and cutting parameters of rotating needles for core biopsy.

    Science.gov (United States)

    Giovannini, Marco; Ren, Huaqing; Cao, Jian; Ehmann, Kornel

    2018-06-15

    Core needle biopsies are widely adopted medical procedures that consist in the removal of biological tissue to better identify a lesion or an abnormality observed through a physical exam or a radiology scan. These procedures can provide significantly more information than most medical tests and they are usually performed on bone lesions, breast masses, lymph nodes and the prostate. The quality of the samples mainly depends on the forces exerted by the needle during the cutting process. The reduction of these forces is critical to extract high-quality tissue samples. The most critical factors that affect the cutting forces are the geometry of the needle tip and its motion while it is penetrating the tissue. However, optimal needle tip configurations and cutting parameters are not well established for rotating insertions. In this paper, the geometry and cutting forces of hollow needles are investigated. The fundamental goal of this study is to provide a series of guidelines for clinicians and surgeons to properly select the optimal tip geometries and speeds. Analytical models related to the cutting angles of several needle tip designs are presented and compared. Several needle tip geometries were manufactured from a 14-gauge cannula, commonly adopted during breast biopsies. The needles were then tested at different speeds and on different phantom tissues. According to these experimental measurements recommendations were formulated for rotating needle insertions. The findings of this study can be applied and extended to several biopsy procedures in which a cannula is used to extract tissue samples. Copyright © 2018 Elsevier Ltd. All rights reserved.

  12. Design and realization of the waster level system based on PLC and configuration technology

    International Nuclear Information System (INIS)

    Hu Kaiming; Ge Yuanxiang; Fu Zhijian

    2012-01-01

    The process control experimental system takes the water level as the controlled object. In this system PLC is adopted as a controller and the electric valve for the actuator. Based on the water level system in the establishment of the mathematical model, the system uses PID algorithm, the PID parameters are tuning, MCGS configuration environment is built and debugs in the environment. Experiments show that the system is running smoothly, the overshoot is 18%, the setting time is about 3 min. The control indicators have reached the engineering requirements and the control effect is good. (authors)

  13. Design considerations in mechanical face seals for improved performance. I - Basic configurations

    Science.gov (United States)

    Ludwig, L. P.; Greiner, H. F.

    1977-01-01

    Basic assembly configurations of the mechanical face seal are described and some advantages associated with each are listed. The various forms of seal components (the primary seal, secondary seal, etc.) are illustrated, and functions pointed out. The technique of seal pressure balancing and its application is described; and the concept of the PV factor, its different forms and limitations are discussed. Brief attention is given to seal lubrication since it is covered in detail in a companion paper. Finally, the operating conditions for various applications of low pressure seals (aircraft transmissions) are listed, and the seal failure mode of a particular application is discussed.

  14. Design considerations in mechanical face seals for improved performance. 1: Basic configurations

    Science.gov (United States)

    Ludwig, L. P.; Greiner, H. F.

    1977-01-01

    Basic assembly configurations of the mechanical face seal are described and some advantages associated with each are listed. The various forms of seal components are illustrated, and functions pointed out. The technique of seal pressure balancing and its application are described; and the concept of the PV factor, its different forms and limitations are discussed. Brief attention is given to seal lubrication since it is covered in detail in a companion paper. Finally, the operating conditions for various applications of low pressure seals (aircraft transmissions) are listed, and the seal failure mode of a particular application is discussed.

  15. Design of an Oximeter Based on LED-LED Configuration and FPGA Technology

    OpenAIRE

    Stojanovic, Radovan; Karadaglic, Dejan

    2013-01-01

    A fully digital photoplethysmographic (PPG) sensor and actuator has been developed. The sensing circuit uses one Light Emitting Diode (LED) for emitting light into human tissue and one LED for detecting the reflectance light from human tissue. A Field Programmable Gate Array (FPGA) is used to control the LEDs and determine the PPG and Blood Oxygen Saturation (SpO2). The configurations with two LEDs and four LEDs are developed for measuring PPG signal and Blood Oxygen Saturation (SpO2). N-LEDs...

  16. Stress analysis of the conceptual design configurations of constant tension D-shaped superconducting toroidal field coils for TNS

    International Nuclear Information System (INIS)

    Fernades, R.; Smith, R.A.

    1977-01-01

    Conceptual design configurations of D-shaped toroidal field coils applicable to the TNS program are studied under the action of the toroidal field loading condition and the vertical field loading condition, but not the fault condition. Although the analysis is specific to an 8 Tesla design using a niobium titanium superconductor, the results can be extended to a coil with a different conductor material and subjected to a field of different magnitude provided the condition of linear elasticity is not violated. The analysis technique used is the finite element method, with three dimensional finite elements defined in the ANSYS computer code, and supplemented by closed form analytical solutions

  17. Design Optimization of Long-Span Cold-Formed Steel Portal Frames Accounting for Effect of Knee Brace Joint Configuration

    OpenAIRE

    Thanh Duoc Phan; James B. P. Lim; Meheron Selowara Joo; Hieng-Ho Lau

    2017-01-01

    The application of cold-formed steel channel sections for portal frames becomes more popular for industrial and residential purposes. Experimental tests showed that such structures with long-span up to 20 m can be achieved when knee brace joints are included. In this paper, the influence of knee brace configuration on the optimum design of long-span cold-formed steel portal frames is investigated. The cold-formed steel portal frames are designed using Eurocode 3 under ultimate limit states. A...

  18. Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

    2000-01-01

    The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium

  19. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  20. Design of a Modular E-Core Flux Concentrating Axial Flux Machine: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Husain, Tausif; Sozer, Yilmaz; Husain, Iqbal; Muljadi, Eduard

    2015-08-24

    In this paper a novel E-Core axial flux machine is proposed. The machine has a double-stator, single-rotor configuration with flux-concentrating ferrite magnets and pole windings across each leg of an E-Core stator. E-Core stators with the proposed flux-concentrating rotor arrangement result in better magnet utilization and higher torque density. The machine also has a modular structure facilitating simpler construction. This paper presents a single-phase and a three-phase version of the E-Core machine. Case studies for a 1.1-kW, 400-rpm machine for both the single-phase and three-phase axial flux machines are presented. The results are verified through 3D finite element analysis. facilitating simpler construction. This paper presents a single-phase and a three-phase version of the E-Core machine. Case studies for a 1.1-kW, 400-rpm machine for both the single-phase and three-phase axial flux machines are presented. The results are verified through 3D finite element analysis.

  1. Observation of inward and outward particle convection in the core of electron cyclotron heated and current driven plasmas in the Tokamak a Configuration Variable

    International Nuclear Information System (INIS)

    Furno, I.; Weisen, H.

    2003-01-01

    In the Tokamak a Configuration Variable [F. Hofmann, J.B. Lister, M. Anton et al., Plasma Phys. Controlled Fusion 36, B277 (1994)], inward or outward convection in the core of electron cyclotron heated and current driven plasmas is observed, depending on discharge conditions. In sawtoothing discharges with central electron cyclotron heating, outward convection is observed when a quasicontinuous m=1 kink mode is present, resulting in inverted sawteeth on the central electron density, while in the absence thereof, inward convection between successive sawtooth crashes leads to 'normal' sawteeth. The occurrence of a kink mode depends sensitively on plasma triangularity. When sawteeth are stabilized with central co- or counterelectron cyclotron current drive, stationary hollow electron density profiles are observed in the presence of m=1 modes, while peaked or flat profiles are observed in magnetohydrodynamic quiescent discharges. The observation of peaked density profiles in fully electron cyclotron driven plasmas demonstrates that pinch processes other than the Ware pinch must be responsible for these phenomena

  2. Design Optimization of Long-Span Cold-Formed Steel Portal Frames Accounting for Effect of Knee Brace Joint Configuration

    Directory of Open Access Journals (Sweden)

    Thanh Duoc Phan

    2017-12-01

    Full Text Available The application of cold-formed steel channel sections for portal frames becomes more popular for industrial and residential purposes. Experimental tests showed that such structures with long-span up to 20 m can be achieved when knee brace joints are included. In this paper, the influence of knee brace configuration on the optimum design of long-span cold-formed steel portal frames is investigated. The cold-formed steel portal frames are designed using Eurocode 3 under ultimate limit states. A novel method in handling design constraints integrated with genetic algorithm is proposed for searching the optimum design of cold-formed steel portal frames. The result showed that the proposed routine for design optimization effectively searched the near global optimum solution with the computational time is approximate 50% faster than methods being popularly used in literature. The optimum configuration for knee brace joint can reduce the section size of rafter and so the lighter frame could be obtained especially for long-span portal frame. The minimum weight of main frame obtained from optimization process is approximate 19.72% lighter than a Benchmark Frame used in the full-scale experimental test.

  3. Increasing the neutron flux study for the TRR-II core design

    International Nuclear Information System (INIS)

    Chen, C.-H.; Yang, J.-T.; Chou, Y.-C.

    1999-01-01

    The maximum unperturbed thermal flux of the originally proposed core design, which is a 6x6 square arrangement with power level of 20 MW and has been presented at the 6th Meeting of IGORR, for the TRR-II reactor is about 2.0x10 14 n/cm 2 -sec. However, it is no longer satisfied the user's requirement, that is, it must reach at least 2.5x10 14 n/cm 2 -sec. In order to enhance the thermal neutron flux, one of the most effective ways is to increase the average power density. Therefore, two new designs with more compact cores are then proposed and studied. One is 5x6 rectangular arrangement with power of 20 MW; the other one is 5x5 square arrangement with power of 16 MW. It is for sure that both core designs can satisfy thermal hydraulic safety limits. The designed parameters related to neutronics are listed and compared fundamentally. According to our calculation, although both cores have similar average power density, the results show that the 5x6/20 MW design has the maximum unperturbed thermal flux in the D 2 O region about 2.7x10 14 n/cm 2 -sec, and the 5x5/16 MW design has 2.5x10 14 n/cm 2 -sec. The maximum thermal flux in the neighborhood of the longer side of the 5x6 core is about 7% higher than the one in the neighborhood of any side of the 5x5 core. This 'long-side effect' gives the 5x6/20 MW core design an advantage of the utilization of the thermal neutron flux in the D 2 O region. In addition, the 5x5 core is also more sensitive to the reactivity change on account of in-core irradiation test facilities. Therefore, under overall considerations the 5x6/20 MW core design is chosen for further detailed design. (author)

  4. Configuring the User as Everybody: Gender and Design in Information and Communication Technologies

    NARCIS (Netherlands)

    Oudshoorn, Nelly E.J.; Rommes, E.W.M.; Stienstra, Marcelle

    2004-01-01

    Based on two case studies of the design of electronic communication networks developed in the public and private sector, this article explores the barriers within current design cultures to account for the needs and diversity of users. Whereas the constraints on user-centered design are usually

  5. Configuring the User as Everybody. Gender and Design Cultures in Information and Communication Technologies

    NARCIS (Netherlands)

    Oudshoorn, N.E.J.; Rommes, E.W.M.; Stienstra, M.

    2004-01-01

    Based on two case studies of the design of electronic communication networks developed in the public and private sector, this article explores the barriers within current design cultures to account for the needs and diversity of users. Whereas the constraints on user-centered design are usually

  6. Exploring cosmic origins with CORE: Survey requirements and mission design

    Science.gov (United States)

    Delabrouille, J.; de Bernardis, P.; Bouchet, F. R.; Achúcarro, A.; Ade, P. A. R.; Allison, R.; Arroja, F.; Artal, E.; Ashdown, M.; Baccigalupi, C.; Ballardini, M.; Banday, A. J.; Banerji, R.; Barbosa, D.; Bartlett, J.; Bartolo, N.; Basak, S.; Baselmans, J. J. A.; Basu, K.; Battistelli, E. S.; Battye, R.; Baumann, D.; Benoít, A.; Bersanelli, M.; Bideaud, A.; Biesiada, M.; Bilicki, M.; Bonaldi, A.; Bonato, M.; Borrill, J.; Boulanger, F.; Brinckmann, T.; Brown, M. L.; Bucher, M.; Burigana, C.; Buzzelli, A.; Cabass, G.; Cai, Z.-Y.; Calvo, M.; Caputo, A.; Carvalho, C.-S.; Casas, F. J.; Castellano, G.; Catalano, A.; Challinor, A.; Charles, I.; Chluba, J.; Clements, D. L.; Clesse, S.; Colafrancesco, S.; Colantoni, I.; Contreras, D.; Coppolecchia, A.; Crook, M.; D'Alessandro, G.; D'Amico, G.; da Silva, A.; de Avillez, M.; de Gasperis, G.; De Petris, M.; de Zotti, G.; Danese, L.; Désert, F.-X.; Desjacques, V.; Di Valentino, E.; Dickinson, C.; Diego, J. M.; Doyle, S.; Durrer, R.; Dvorkin, C.; Eriksen, H. K.; Errard, J.; Feeney, S.; Fernández-Cobos, R.; Finelli, F.; Forastieri, F.; Franceschet, C.; Fuskeland, U.; Galli, S.; Génova-Santos, R. T.; Gerbino, M.; Giusarma, E.; Gomez, A.; González-Nuevo, J.; Grandis, S.; Greenslade, J.; Goupy, J.; Hagstotz, S.; Hanany, S.; Handley, W.; Henrot-Versillé, S.; Hernández-Monteagudo, C.; Hervias-Caimapo, C.; Hills, M.; Hindmarsh, M.; Hivon, E.; Hoang, D. T.; Hooper, D. C.; Hu, B.; Keihänen, E.; Keskitalo, R.; Kiiveri, K.; Kisner, T.; Kitching, T.; Kunz, M.; Kurki-Suonio, H.; Lagache, G.; Lamagna, L.; Lapi, A.; Lasenby, A.; Lattanzi, M.; Le Brun, A. M. C.; Lesgourgues, J.; Liguori, M.; Lindholm, V.; Lizarraga, J.; Luzzi, G.; Macìas-P{érez, J. F.; Maffei, B.; Mandolesi, N.; Martin, S.; Martinez-Gonzalez, E.; Martins, C. J. A. P.; Masi, S.; Massardi, M.; Matarrese, S.; Mazzotta, P.; McCarthy, D.; Melchiorri, A.; Melin, J.-B.; Mennella, A.; Mohr, J.; Molinari, D.; Monfardini, A.; Montier, L.; Natoli, P.; Negrello, M.; Notari, A.; Noviello, F.; Oppizzi, F.; O'Sullivan, C.; Pagano, L.; Paiella, A.; Pajer, E.; Paoletti, D.; Paradiso, S.; Partridge, R. B.; Patanchon, G.; Patil, S. P.; Perdereau, O.; Piacentini, F.; Piat, M.; Pisano, G.; Polastri, L.; Polenta, G.; Pollo, A.; Ponthieu, N.; Poulin, V.; Prêle, D.; Quartin, M.; Ravenni, A.; Remazeilles, M.; Renzi, A.; Ringeval, C.; Roest, D.; Roman, M.; Roukema, B. F.; Rubiño-Martin, J.-A.; Salvati, L.; Scott, D.; Serjeant, S.; Signorelli, G.; Starobinsky, A. A.; Sunyaev, R.; Tan, C. Y.; Tartari, A.; Tasinato, G.; Toffolatti, L.; Tomasi, M.; Torrado, J.; Tramonte, D.; Trappe, N.; Triqueneaux, S.; Tristram, M.; Trombetti, T.; Tucci, M.; Tucker, C.; Urrestilla, J.; Väliviita, J.; Van de Weygaert, R.; Van Tent, B.; Vennin, V.; Verde, L.; Vermeulen, G.; Vielva, P.; Vittorio, N.; Voisin, F.; Wallis, C.; Wandelt, B.; Wehus, I. K.; Weller, J.; Young, K.; Zannoni, M.

    2018-04-01

    Future observations of cosmic microwave background (CMB) polarisation have the potential to answer some of the most fundamental questions of modern physics and cosmology, including: what physical process gave birth to the Universe we see today? What are the dark matter and dark energy that seem to constitute 95% of the energy density of the Universe? Do we need extensions to the standard model of particle physics and fundamental interactions? Is the ΛCDM cosmological scenario correct, or are we missing an essential piece of the puzzle? In this paper, we list the requirements for a future CMB polarisation survey addressing these scientific objectives, and discuss the design drivers of the COREmfive space mission proposed to ESA in answer to the "M5" call for a medium-sized mission. The rationale and options, and the methodologies used to assess the mission's performance, are of interest to other future CMB mission design studies. COREmfive has 19 frequency channels, distributed over a broad frequency range, spanning the 60–600 GHz interval, to control astrophysical foreground emission. The angular resolution ranges from 2' to 18', and the aggregate CMB sensitivity is about 2 μKṡarcmin. The observations are made with a single integrated focal-plane instrument, consisting of an array of 2100 cryogenically-cooled, linearly-polarised detectors at the focus of a 1.2-m aperture cross-Dragone telescope. The mission is designed to minimise all sources of systematic effects, which must be controlled so that no more than 10‑4 of the intensity leaks into polarisation maps, and no more than about 1% of E-type polarisation leaks into B-type modes. COREmfive observes the sky from a large Lissajous orbit around the Sun-Earth L2 point on an orbit that offers stable observing conditions and avoids contamination from sidelobe pick-up of stray radiation originating from the Sun, Earth, and Moon. The entire sky is observed repeatedly during four years of continuous scanning

  7. Simulation of the Long period Core Design for WH type of KHNP

    International Nuclear Information System (INIS)

    Jung, Ji-Eun; Moon, Sang-Rae

    2016-01-01

    The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve

  8. Simulation of the Long period Core Design for WH type of KHNP

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ji-Eun; Moon, Sang-Rae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve.

  9. PIV Logon Configuration Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Glen Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-04

    This document details the configurations and enhancements implemented to support the usage of federal Personal Identity Verification (PIV) Card for logon on unclassified networks. The guidance is a reference implementation of the configurations and enhancements deployed at the Los Alamos National Laboratory (LANL) by Network and Infrastructure Engineering – Core Services (NIE-CS).

  10. Recriticality, a Key Phenomenon to Investigate in Core Disruptive Accident Scenarios of Current and Future Fast Reactor Designs

    International Nuclear Information System (INIS)

    Maschek, W.; Rineiski, A.; Flad, M.; Kriventsev, V.; Gabrielli, F.; Morita, K.

    2012-01-01

    Final comments and conclusions: • Modern plants, should have performed better under Fukushima type event. • In future fast reactor systems significantly higher active and passive safety features are installed, which should cope with events like Fukushima. • One important lesson: put a focus on rare initiators, accident routes and consequences that are neither expected nor have been observed, events that are categorized under ‘black swans’. • Importance of severe accident research demonstrated - both analytically and experimentally for assessing and interpreting accident scenarios and developments. Precondition for developing preventive & mitigative safety measures. Passive safety measures are in the focus of advanced design options and must work under conditions of multiple loads and aggravating events. • Fast reactor systems behavior as the SFR under severe accident conditions: – In fast spectrum systems as the SFR the core is not in its neutronically most reactive configuration and SFRs may be loaded with MAs for waste management; – Recriticalities have a high probability because of the higher enrichment levels; – Short time scales have to be envisioned for core melt-down; – Decay heat levels might be significantly higher, if MA bearing fuel is involved. • Improve design by measures for prevention and/or mitigation of recriticalities; – High reliability of simulations required for proof; • Assessment of fuel relocated on peripheral structures; • Preventive/mitigating measures should not replace containment measures

  11. Design and field configuration for a 14.4 GHz ECR ion source in Kolkata

    International Nuclear Information System (INIS)

    Rashid, M.H.; Bose, D.K.; Mallik, C.; Bhandari, R.K.

    2001-01-01

    The K500 cyclotron under construction will be capable of accelerating ions like O 6+ , Ne 4+ , Ar 16+ , Kr 27+ etc. We aim to get ∼200 euA maximum intensity of the extracted beam of O 6+ from the ion source and decided to have >2B ECR magnetic field on the cylindrical surface and the injection ends of the plasma chamber (P Ch) and slightly less than this at the extraction end. The success of the high field operation of ECRs at other places (U-AECR at LBL) suggests generation of proper magnetic field configuration for the 14.4 GHz microwave heating. The absolute composite magnetic field have been evaluated due to the coils (C1,C2) at the two ends and a -ve coil (NC) at the mid-length and a Halbach type sextupole (PM-Hex)

  12. Software system development of NPP plant DiD risk monitor. Basic design of software configuration

    International Nuclear Information System (INIS)

    Yoshikawa, Hidekazu; Nakagawa, Takashi

    2015-01-01

    A new risk monitor system is under development which can be applied not only to prevent severe accident in daily operation but also to serve as to mitigate the radiological hazard just after severe accident happens and long term management of post-severe accident consequences. The fundamental method for the new risk monitor system is first given on how to configure the Plant Defense in-Depth (DiD) Risk Monitor by object-oriented software system based on functional modeling approach. In this paper, software system for the plant DiD risk monitor is newly developed by object oriented method utilizing Unified Modeling Language (UML). Usage of the developed DiD risk monitor is also introduced by showing examples for LOCA case of AP1000. (author)

  13. Core designs for new VVER reactors and operational experience of immediate prototypes

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Mokhov, V.; Ryzhov, S.

    2011-01-01

    The paper covers the recent improvements analyzed in order to implement the enhanced core performances. AES-2006 reactor core design is considered from the point of view of its application and improvement in the planned VVER-TOI project and of the possibilities of using the basic engineering solutions for the cores with spectral control. The discussion of several types of mixing grids considered in the paper involves a preliminary assessment of their efficiency and the information on their introduction into pilot operation at the VVER-1000 Units. Special attention is given to the results of the operation of immediate prototypes (TVS-2 and TVS-2M) that corroborate the reliability of the design both with regard for the core geometrical stability and fuel cladding tightness

  14. Design/Operations review of core sampling trucks and associated equipment

    International Nuclear Information System (INIS)

    Shrivastava, H.P.

    1996-01-01

    A systematic review of the design and operations of the core sampling trucks was commissioned by Characterization Equipment Engineering of the Westinghouse Hanford Company in October 1995. The review team reviewed the design documents, specifications, operating procedure, training manuals and safety analysis reports. The review process, findings and corrective actions are summarized in this supporting document

  15. Characteristic features of the core design of high-temperature reactors

    International Nuclear Information System (INIS)

    Brandes, S.; Lohnert, G.

    1975-01-01

    Following a survey on the possible applications of the HTGR depending on the height of the gas exiting temperatures, the core design for both of the fuel element concepts 'sphere' and 'block' is dealt with. The particularities arising from the multiple refueling and the one-way fueling in the design for spherical fuel elements are discussed. (UA/LH) [de

  16. The design of an asynchronous Tiny RISC TM/TR4101 microprocessor core

    DEFF Research Database (Denmark)

    Christensen, Kåre Tais; Jensen, P.; Korger, P.

    1998-01-01

    This paper presents the design of an asynchronous version of the TR4101 embedded microprocessor core developed by LSI Logic Inc. The asynchronous processor, called ARISC, was designed using the same CAD tools and the same standard cell library that was used to implement the TR4101. The paper repo...

  17. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  18. The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

    International Nuclear Information System (INIS)

    Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

    2009-01-01

    The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

  19. Fuel and Core Design Verification for Extended Power Up-rate in Ringhals Unit 3

    International Nuclear Information System (INIS)

    Gabrielsson, Petter; Stepniewski, Marek; Almberger, Jan

    2006-01-01

    Vattenfall's Westinghouse 3-loop PWR Ringhals 3 at the western coast of Sweden is scheduled for an extended power up-rate from 2783 to 3160 MWt in 2007, in the frame of the so called GREAT-project. The project will realize an up-rating initially planned and analysed back in 1995, but with a number of significant improvements outlined in this paper. For the licensing of the up-rated power level, a complete revision of the safety analyses, radiological analyses and systems verifications in FSAR is being performed by Westinghouse Electrics Belgium. The work is performed in close cooperation with Vattenfall in the areas of core calculations and input data. For more than a decade, Vattenfall has performed all core design and reload safety evaluations (RSE) for Ringhals, independent of fuel vendors and safety analysts. In GREAT all core parameters in the safety analysis checklist (SAC) used for the safety analyses are determined based upon a set of nine reference loading patterns designed by Vattenfall covering a wide range of fuel and core designs and extreme cycle-to-cycle variations. To facilitate the calculation of SAC parameters Westinghouse has provided a Reload Safety Evaluation Procedure report (RSEP) with detailed specifications for the calculation of all core parameters used in the analyses. The procedure has been automatized by Vattenfall in a set of scripts executing 3D core simulator calculations and extracting the key results. The same tools will be used in Vattenfall's future RSE for Ringhals 3. This approach is taken to obtain consistency between core designs and core calculations for the safety analyses and the cycle specific calculations, to minimize the risk for future violations of the safety analyses. (authors)

  20. RotCFD: A Viscous Design Tool for Advanced Configurations, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — During initial design studies, parametric variation of vehicle geometry is routine. In addition, rotorcraft engineers traditionally use the wind tunnel to evaluate...

  1. Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

    International Nuclear Information System (INIS)

    Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

    1988-04-01

    As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

  2. Configuration management

    International Nuclear Information System (INIS)

    Beavers, R.R.; Sumiec, K.F.

    1989-01-01

    Increasing regulatory and industry attention has been focused on properly controlling electrical design changes. These changes can be controlled by using configuration management techniques. Typically, there are ongoing modifications to various process systems or additions due to new requirements at every power plant. Proper control of these changes requires that an organized method be used to ensure that all important parameters of the electrical auxiliary systems are analyzed and that these parameters are evaluated accurately. This process, commonly referred to as configuration management, is becoming more important on both fossil and nuclear plants. Recent NRC- and utility-initiated inspections have identified problems due to incomplete analysis of changes to electrical auxiliary systems at nuclear stations

  3. Development of core design/analysis technology for integral reactor; verification of SMART nuclear design by Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)

    2002-03-01

    The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)

  4. Rational Design of Molecular Hole-Transporting Materials for Perovskite Solar Cells: Direct versus Inverted Device Configurations.

    Science.gov (United States)

    Grisorio, Roberto; Iacobellis, Rosabianca; Listorti, Andrea; De Marco, Luisa; Cipolla, Maria Pia; Manca, Michele; Rizzo, Aurora; Abate, Antonio; Gigli, Giuseppe; Suranna, Gian Paolo

    2017-07-26

    Due to a still limited understanding of the reasons making 2,2',7,7'-tetrakis(N,N-di-p-methoxyphenylamine)-9,9'-spirobifluorene (Spiro-OMeTAD) the state-of-the-art hole-transporting material (HTM) for emerging photovoltaic applications, the molecular tailoring of organic components for perovskite solar cells (PSCs) lacks in solid design criteria. Charge delocalization in radical cationic states can undoubtedly be considered as one of the essential prerequisites for an HTM, but this aspect has been investigated to a relatively minor extent. In marked contrast with the 3-D structure of Spiro-OMeTAD, truxene-based HTMs Trux1 and Trux2 have been employed for the first time in PSCs fabricated with a direct (n-i-p) or inverted (p-i-n) architecture, exhibiting a peculiar behavior with respect to the referential HTM. Notwithstanding the efficient hole extraction from the perovskite layer exhibited by Trux1 and Trux2 in direct configuration devices, their photovoltaic performances were detrimentally affected by their poor hole transport. Conversely, an outstanding improvement of the photovoltaic performances in dopant-free inverted configuration devices compared to Spiro-OMeTAD was recorded, ascribable to the use of thinner HTM layers. The rationalization of the photovoltaic performances exhibited by different configuration devices discussed in this paper can provide new and unexpected prospects for engineering the interface between the active layer of perovskite-based solar cells and the hole transporters.

  5. 1995 NASA High-Speed Research Program Sonic Boom Workshop. Volume 2; Configuration Design, Analysis, and Testing

    Science.gov (United States)

    Baize, Daniel G. (Editor)

    1999-01-01

    The High-Speed Research Program and NASA Langley Research Center sponsored the NASA High-Speed Research Program Sonic Boom Workshop on September 12-13, 1995. The workshop was designed to bring together NASAs scientists and engineers and their counterparts in industry, other Government agencies, and academia working together in the sonic boom element of NASAs High-Speed Research Program. Specific objectives of this workshop were to: (1) report the progress and status of research in sonic boom propagation, acceptability, and design; (2) promote and disseminate this technology within the appropriate technical communities; (3) help promote synergy among the scientists working in the Program; and (4) identify technology pacing, the development C, of viable reduced-boom High-Speed Civil Transport concepts. The Workshop was organized in four sessions: Sessions 1 Sonic Boom Propagation (Theoretical); Session 2 Sonic Boom Propagation (Experimental); Session 3 Acceptability Studies-Human and Animal; and Session 4 - Configuration Design, Analysis, and Testing.

  6. The effect of the configuration and the interior design of a virtual weightless space station on human spatial orientation

    Science.gov (United States)

    Aoki, Hirofumi; Ohno, Ryuzo; Yamaguchi, Takao

    2005-05-01

    In a virtual weightless environment, subjects' orientation skills were studied to examine what kind of cognitive errors people make when they moved through the interior space of virtual space stations and what kind of visual information effectively decreases those errors. Subjects wearing a head-mounted display moved from one end to the other end in space station-like routes constructed of rectangular and cubical modules, and did Pointing and Modeling tasks. In Experiment 1, configurations of the routes were changed with such variables as the number of bends, the number of embedding planes, and the number of planes with respect to the body posture. The results indicated that spatial orientation ability was relevant to the variables and that orientational errors were explained by two causes. One of these was that the place, the direction, and the sequence of turns were incorrect. The other was that subjects did not recognize the rotation of the frame of reference, especially when they turned in pitch direction rather than in yaw. In Experiment 2, the effect of the interior design was examined by testing three design settings. Wall colors that showed the allocentric frame of reference and the different interior design of vertical and horizontal modules were effective; however, there was a limit to the effectiveness in complicated configurations.

  7. The effect of the configuration and the interior design of a virtual weightless space station on human spatial orientation.

    Science.gov (United States)

    Aoki, Hirofumi; Ohno, Ryuzo; Yamaguchi, Takao

    2005-01-01

    In a virtual weightless environment, subjects' orientation skills were studied to examine what kind of cognitive errors people make when they moved through the interior space of virtual space stations and what kind of visual information effectively decreases those errors. Subjects wearing a head-mounted display moved from one end to the other end in space station-like routes constructed of rectangular and cubical modules, and did Pointing and Modeling tasks. In Experiment 1, configurations of the routes were changed with such variables as the number of bends, the number of embedding planes, and the number of planes with respect to the body posture. The results indicated that spatial orientation ability was relevant to the variables and that orientational errors were explained by two causes. One of these was that the place, the direction, and the sequence of turns were incorrect. The other was that subjects did not recognize the rotation of the frame of reference, especially when they turned in pitch direction rather than in yaw. In Experiment 2, the effect of the interior design was examined by testing three design settings. Wall colors that showed the allocentric frame of reference and the different interior design of vertical and horizontal modules were effective; however, there was a limit to the effectiveness in complicated configurations. c2005 Published by Elsevier Ltd.

  8. Optimizing a three-element core design for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    West, C.D.

    1995-01-01

    Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core

  9. High Level Analysis, Design and Validation of Distributed Mobile Systems with CoreASM

    Science.gov (United States)

    Farahbod, R.; Glässer, U.; Jackson, P. J.; Vajihollahi, M.

    System design is a creative activity calling for abstract models that facilitate reasoning about the key system attributes (desired requirements and resulting properties) so as to ensure these attributes are properly established prior to actually building a system. We explore here the practical side of using the abstract state machine (ASM) formalism in combination with the CoreASM open source tool environment for high-level design and experimental validation of complex distributed systems. Emphasizing the early phases of the design process, a guiding principle is to support freedom of experimentation by minimizing the need for encoding. CoreASM has been developed and tested building on a broad scope of applications, spanning computational criminology, maritime surveillance and situation analysis. We critically reexamine here the CoreASM project in light of three different application scenarios.

  10. Design and safety studies on an EFIT core with CERMET fuel

    International Nuclear Information System (INIS)

    Chen, Xue-Nong; Rineiski, Andrei; Liu, Ping; Maschek, Werner; Matzerath Boccaccini, Claudia; Gabrielli, Fabrizio; Sobolev, Vitaly

    2008-01-01

    Within the EUROTRANS Programme a European Facility for Industrial Transmutation (EFIT) is under development. This paper deals with the design and safety analyses of an EFIT core with Mo-matrix based CERMET fuel. A three zone core design was developed, which satisfies the EFIT general and specific requirements. The fuel/matrix ratio in each zone is determined for a suitable subcritical level at a k eff of about 0.97 and a total form factor around 1.5. The Pu/MA ratio also determines the transmutation rate and the burn-up characteristics, ranging between 46/54 at% to 40/60 at% for optimizing the reactivity swing and the MA transmutation efficiency. Based on the preliminary core design, safety calculations are performed with SIMMER-III for three types of transient: the unprotected loss of flow (ULOF), the unprotected transient of over power (UTOP) and the unprotected blockage accident (UBA). It can be shown that in the CERMET core the fuel and clad design limits are not violated under the conditions of ULOF and UTOP. In the UBA case, pin failures will happen and lead to a local voiding and reactivity insertion, but a fuel sweep-out process leads to a power reduction and restricts the core degradation. (authors)

  11. Optimization of core reload design for low leakage fuel management in pressurized water reactors

    International Nuclear Information System (INIS)

    Kim, Y.J.

    1986-01-01

    A new method was developed to optimize pressurized water reactor core reload design for low leakage fuel management, a strategy recently adopted by most utilities to extend cycle length and mitigate pressurized thermal shock concerns. The method consists of a two-stage optimization process which provides the maximum cycle length for a given fresh fuel loading subject to power peaking constraints. In the first stage, a best fuel arrangement is determined at the end of cycle in the absence of burnable poisons. A direct search method is employed in conjunction with a constant power, Haling depletion. In the second stage, the core control poison requirements are determined using a linear programming technique. The solution provides the fresh fuel burnable poison loading required to meet core power peaking constraints. An accurate method of explicitly modeling burnable absorbers was developed for this purpose. The design method developed here was implemented in a currently recognized fuel licensing code, SIMULATE, that was adapted to the CYBER-205 computer. This methodology was applied to core reload design of cycles 9 and 10 for the Commonwealth Edison Zion, Unit-1 Reactor. The results showed that the optimum loading pattern for cycle 9 yielded almost a 9% increase in the cycle length while reducing core vessel fluence by 30% compared with the reference design used by Commonwealth Edison

  12. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    International Nuclear Information System (INIS)

    Bi, G.; Liu, C.; Si, S.

    2012-01-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

  13. Study of advanced composite structural design concepts for an arrow wing supersonic cruise configuration

    Science.gov (United States)

    Turner, M. J.; Grande, D. L.

    1978-01-01

    Based on estimated graphite and boron fiber properties, allowable stresses and strains were established for advanced composite materials. Stiffened panel and conventional sandwich panel concepts were designed and analyzed, using graphite/polyimide and boron/polyimide materials. The conventional sandwich panel was elected as the structural concept for the modified wing structure. Upper and lower surface panels of the arrow wing structure were then redesigned, using high strength graphite/polyimide sandwich panels, retaining the titanium spars and ribs from the prior study. The ATLAS integrated analysis and design system was used for stress analysis and automated resizing of surface panels. Flutter analysis of the hybrid structure showed a significant decrease in flutter speed relative to the titanium wing design. The flutter speed was increased to that of the titanium design by selective increase in laminate thickness and by using graphite fibers with properties intermediate between high strength and high modulus values.

  14. RotCFD: A Viscous Design Tool for Advanced Configurations, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The incorporation of viscous analysis in design is vital for a complete understanding of aerodynamic problems. This proposal offers to develop and integrate with...

  15. Design and development of a run-time monitor for multi-core architectures in cloud computing.

    Science.gov (United States)

    Kang, Mikyung; Kang, Dong-In; Crago, Stephen P; Park, Gyung-Leen; Lee, Junghoon

    2011-01-01

    Cloud computing is a new information technology trend that moves computing and data away from desktops and portable PCs into large data centers. The basic principle of cloud computing is to deliver applications as services over the Internet as well as infrastructure. A cloud is a type of parallel and distributed system consisting of a collection of inter-connected and virtualized computers that are dynamically provisioned and presented as one or more unified computing resources. The large-scale distributed applications on a cloud require adaptive service-based software, which has the capability of monitoring system status changes, analyzing the monitored information, and adapting its service configuration while considering tradeoffs among multiple QoS features simultaneously. In this paper, we design and develop a Run-Time Monitor (RTM) which is a system software to monitor the application behavior at run-time, analyze the collected information, and optimize cloud computing resources for multi-core architectures. RTM monitors application software through library instrumentation as well as underlying hardware through a performance counter optimizing its computing configuration based on the analyzed data.

  16. Design and Development of a Run-Time Monitor for Multi-Core Architectures in Cloud Computing

    Directory of Open Access Journals (Sweden)

    Junghoon Lee

    2011-03-01

    Full Text Available Cloud computing is a new information technology trend that moves computing and data away from desktops and portable PCs into large data centers. The basic principle of cloud computing is to deliver applications as services over the Internet as well as infrastructure. A cloud is a type of parallel and distributed system consisting of a collection of inter-connected and virtualized computers that are dynamically provisioned and presented as one or more unified computing resources. The large-scale distributed applications on a cloud require adaptive service-based software, which has the capability of monitoring system status changes, analyzing the monitored information, and adapting its service configuration while considering tradeoffs among multiple QoS features simultaneously. In this paper, we design and develop a Run-Time Monitor (RTM which is a system software to monitor the application behavior at run-time, analyze the collected information, and optimize cloud computing resources for multi-core architectures. RTM monitors application software through library instrumentation as well as underlying hardware through a performance counter optimizing its computing configuration based on the analyzed data.

  17. In core reload design for cycle 4 of Daya Bay nuclear power station both units

    International Nuclear Information System (INIS)

    Zhang Zongyao; Liu Xudong; Xian Chunyu; Li Dongsheng; Zhang Hong; Liu Changwen; Rui Min; Wang Yingming; Zhao Ke; Zhang Hong; Xiao Min

    1998-01-01

    The basic principles and the contents of the reload design for Daya Bay nuclear power station are briefly introduced. The in core reload design results, and the comparison between the calculated values and the measured values of both units the fourth cycle are also given. The reload design results of the two units satisfy all the economic requirements and safety criteria. The experimented results shown that the predicated values are tally good with all the measurement values

  18. Design criteria for a self-actuated shutdown system to ensure limitation of core damage

    International Nuclear Information System (INIS)

    Deane, N.A.; Atcheson, D.B.

    1981-09-01

    Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times

  19. An approach to development of structural design criteria for highly irradiated core components

    International Nuclear Information System (INIS)

    Nelson, D.V.

    1980-01-01

    The advent of the fast breeder reactor presents novel challenges in structural design and materials engineering. For instance, the core components of these reactors experience high energy neutron irradiation at elevated temperature, which causes significant time-dependent changes in material behaviour, such as a progressive loss of ductility. New structural design criteria are needed to extend elevated temperature design-by-analysis to account for these changes. Alloys best able to cope with the demands of the core operating environment are being explored and their structural behaviour characterized. The purpose of this paper is to illustrate an approach used in the development of core component structural design criteria. To do this, several design rules, plus brief rationale, from draft RDT Standards F9-7, -8 and -9 will be presented. These recently completed standards ('Structural Design Guidelines for Breeder Reactor Core Components') were prepared for the U.S. Department of Energy and represent a consensus among most organizations participating in the U.S. breeder program. (author)

  20. Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

    International Nuclear Information System (INIS)

    LECHELT, J.A.

    2000-01-01

    The purpose of the design compliance matrix (DCM) is to provide a single-source document of all design requirements associated with the fifteen subsystems that make up the rotary mode core sampling (RMCS) system. It is intended to be the baseline requirement document for the RMCS system and to be used in governing all future design and design verification activities associated with it. This document is the DCM for the RMCS system used on Hanford single-shell radioactive waste storage tanks. This includes the Exhauster System, Rotary Mode Core Sample Trucks, Universal Sampling System, Diesel Generator System, Distribution Trailer, X-Ray Cart System, Breathing Air Compressor, Nitrogen Supply Trailer, Casks and Cask Truck, Service Trailer, Core Sampling Riser Equipment, Core Sampling Support Trucks, Foot Clamp, Ramps and Platforms and Purged Camera System. Excluded items are tools such as light plants and light stands. Other items such as the breather inlet filter are covered by a different design baseline. In this case, the inlet breather filter is covered by the Tank Farms Design Compliance Matrix