WorldWideScience

Sample records for core nuclear design

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

    Science.gov (United States)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    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.

  2. 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.

  3. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  4. Design and Performance of South Ukraine Nuclear Power Plant Mixed Cores

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Baydulin, V.; Zhukov, A. I.; Latorre, Richard

    2011-09-24

    In 2010, 42 Westinghouse fuel assemblies (WFAs) were loaded into the core of South Ukraine Nuclear Power Plant (SUNPP) Unit 3 after four successful cycles with 6 Westinghouse Lead Test Assemblies. The scope of safety substantiating documents required for the regulatory approval of this mixed core was extended considerably, particularly with development and implementation of new methodologies and 3-D kinetic codes. Additional verification for all employed codes was also performed. Despite the inherent hydraulic non-uniformity of a mixed core, it was possible to demonstrate that all design and operating restrictions for three different types of fuel (TVS-M, TVSA and WFA) loaded in the core were conservatively met. This paper provides the main results from the first year of operation of the core loaded with 42 WFAs, the predicted parameters for the transition and equilibrium cycles with WFAs, comparisons of predicted versus measured core parameters, as well as the acceptable margin evaluation results for reactivity accidents using the 3-D kinetic codes. To date WFA design parameters have been confirmed by operation experience.

  5. Design Optimization of Nuclear Vapor Thermal Rocket Core - A Thermo-Mechanical Study

    Science.gov (United States)

    Keshavmurthy, Shyam P.; Watanabe, Yoichi; Dugan, Edward T.; Diaz, Nils J.

    1994-07-01

    Fuel structural materials for the Nuclear Vapor Thermal Rocket (NVTR) are exposed to very high temperature vapor fuel in the fuel channel and to high temperature but cooler propellant in the coolant channel. This temperature difference leads to thermal stress in the fuel element. There is also a mismatch in the value of coefficients of thermal expansion between the fuel element material and the coating material that could lead to failure of the coating. The stress in the coating and the fuel element material is dependent on the power density of the core and also on the arrangement of fuel and coolant channels. In order to achieve higher power density, the fuel element design has to be optimized to yield lower stress. Analytical studies found that carbon/carbon composite hexagonal fuel elements employing a square lattice arrangement of multiple UF4 fuel and hydrogen coolant channels yield maximum stress intensities well below fuel element materials stress limit.

  6. Cost-based optimization of a nuclear reactor core design: a preliminary model

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F.; Alves Filho, Hermes [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico. Dept. de Modelagem Computacional]. E-mails: wfsacco@iprj.uerj.br; halves@iprj.uerj.br; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil). Div. de Reatores]. E-mail: cmnap@ien.gov.br

    2007-07-01

    A new formulation of a nuclear core design optimization problem is introduced in this article. Originally, the optimization problem consisted in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the radial power peaking factor in a three-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. Here, we address the same problem using the minimization of the fuel and cladding materials costs as the objective function, and the radial power peaking factor as an operational constraint. This cost-based optimization problem is attacked by two metaheuristics, the standard genetic algorithm (SGA), and a recently introduced Metropolis algorithm called the Particle Collision Algorithm (PCA). The two algorithms are submitted to the same computational effort and their results are compared. As the formulation presented is preliminary, more elaborate models are also discussed (author)

  7. CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2015-04-01

    Full Text Available Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR. The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR is upper than 1.25, Hot Excess Reactivity (HOTXS is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant. Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk

  8. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  9. Development of a standard data base for FBR core nuclear design. 8. Compilation of JUPITER analytical results

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Makoto; Sugino, Kazuteru; Yokoyama, Kenji [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Sato, Wakaei; Numata, Kazuyuki; Iwai, Takehiko

    1997-11-01

    A standard data base for LMFBR core nuclear design has been developed to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as demonstration or commercial FBRs. To develop the data base, extensive work has been performed to accumulate and evaluate many kinds of results from fast reactor physics experiments and their analyses. The present report summarizes the analytical results of the JUPITER experiments, using the most recent nuclear data library (JENDL-3.2) and the latest analytical methods in a consistent manner. The total number of JUPITER C/E values obtained here exceeds 2,300, which cover most of the JUPITER data in the nuclear design data base. The analytical results will be combined with the sensitivity coefficients and experimental and analytical error values as a whole, and are expected to contribute the improvement of large FBR core design methods by means of a unified cross-section set for the demonstration FBR and various physical information. (J.P.N.). 236 refs.

  10. Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

    Science.gov (United States)

    Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

    2013-01-01

    Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

  11. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    Science.gov (United States)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  12. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the si

  13. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  14. A Metropolis algorithm combined with Nelder-Mead Simplex applied to nuclear reactor core design

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F. [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil)], E-mail: wfsacco@iprj.uerj.br; Filho, Hermes Alves; Henderson, Nelio [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil); Oliveira, Cassiano R.E. de [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2008-05-15

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder-Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. The new metaheuristic performs better than the genetic algorithm, particle swarm optimization, and the Metropolis algorithms PCA and the Great Deluge Algorithm, thus demonstrating its potential for other applications.

  15. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  16. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  17. 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)

  18. Nuclear design characteristics of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung Chan; Park, Sang Yoon; Lee, Ki Bog; Zee, Sung Quun; Chang, Moon Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Nuclear design bases for System-Integrated Modular Advanced ReacTor(SMART) core are presented. Based on the proposed design bases, a SMART core loading pattern is constructed and its nuclear characteristics are studied. The proposed core loading pattern satisfies 3-year cycle length and soluble boron-free operation requirements at any time during the cycle. 10 refs., 2 figs., 1 tab. (Author)

  19. A feasibility study for the application of enriched gadolinia burnable absorber rods in nuclear core design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung Chan; Zee, Sung Quun; Kim, Kang Seog; Song, Jae Seung

    2000-12-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions.

  20. SMART core protection system design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. K.; Park, H. Y.; Koo, I. S. [KAERI, Taejon (Korea, Republic of); Park, H. S.; Kim, J. S.; Son, C. H. [Samchang Enterprise Co., Ltd., Taejon (Korea, Republic of)

    2003-10-01

    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.

  1. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  2. Nuclear gas core propulsion research program

    Science.gov (United States)

    Diaz, Nils J.; Dugan, Edward T.; Anghaie, Samim

    1993-01-01

    Viewgraphs on the nuclear gas core propulsion research program are presented. The objectives of this research are to develop models and experiments, systems, and fuel elements for advanced nuclear thermal propulsion rockets. The fuel elements under investigation are suitable for gas/vapor and multiphase fuel reactors. Topics covered include advanced nuclear propulsion studies, nuclear vapor thermal rocket (NVTR) studies, and ultrahigh temperature nuclear fuels and materials studies.

  3. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  4. CAC - NUCLEAR THERMAL ROCKET CORE ANALYSIS CODE

    Science.gov (United States)

    Clark, J. S.

    1994-01-01

    One of the most important factors in the development of nuclear rocket engine designs is to be able to accurately predict temperatures and pressures throughout a fission nuclear reactor core with axial hydrogen flow through circular coolant passages. CAC is an analytical prediction program to study the heat transfer and fluid flow characteristics of a circular coolant passage. CAC predicts as a function of time axial and radial fluid conditions, passage wall temperatures, flow rates in each coolant passage, and approximate maximum material temperatures. CAC incorporates the hydrogen properties model STATE to provide fluid-state relations, thermodynamic properties, and transport properties of molecular hydrogen in any fixed ortho-para combination. The program requires the general core geometry, the core material properties as a function of temperature, the core power profile, and the core inlet conditions as function of time. Although CAC was originally developed in FORTRAN IV for use on an IBM 7094, this version is written in ANSI standard FORTRAN 77 and is designed to be machine independent. It has been successfully compiled on IBM PC series and compatible computers running MS-DOS with Lahey F77L, a Sun4 series computer running SunOS 4.1.1, and a VAX series computer running VMS 5.4-3. CAC requires 300K of RAM under MS-DOS, 422K of RAM under SunOS, and 220K of RAM under VMS. No sample executable is provided on the distribution medium. Sample input and output data are included. The standard distribution medium for this program is a 5.25 inch 360K MS-DOS format diskette. CAC was developed in 1966, and this machine independent version was released in 1992. IBM-PC and IBM are registered trademarks of International Business Machines. Lahey F77L is a registered trademark of Lahey Computer Systems, Inc. SunOS is a trademark of Sun Microsystems, Inc. VMS is a trademark of Digital Equipment Corporation. MS-DOS is a registered trademark of Microsoft Corporation.

  5. 堆芯核设计程序CYCAS动力学模型开发%Development of Kinetics Model in Core Nuclear Design Code CYCAS

    Institute of Scientific and Technical Information of China (English)

    毕光文; 汤春桃; 杨波

    2016-01-01

    The kinetics model and its numerical verification were studied for core nuclear design code CYCAS .The kinetics model employed by CYCAS code was introduced in detail .In order to verify the effectiveness of the kinetics model , the L M W transient benchmark and the dynamic insertion issue of control rod in AP1000 core were simulated and analyzed .The calculation results show that the kinetics model of CYCAS code could obtain reliable results .%对堆芯核设计程序CYCAS的动力学模型及其数值验证进行了研究.详细介绍了CYCAS程序采用的动力学模型.为验证模型的有效性,对L M W瞬态基准题和基于AP1000堆芯动态插棒问题进行了数值模拟和分析.结果表明,CYCAS程序的动力学模型可获得可靠的计算结果.

  6. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  7. Wire core reactor for nuclear thermal propulsion

    Science.gov (United States)

    Harty, Richard B.; Brengle, Robert G.

    1993-01-01

    Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over alternate layers of spacer wires, which forms a rugged annular lattice. Hydrogen flow in the core is annular, flowing from inside to outside. In addition to the concepts compact size and good heat transfer, the core has excellent power-flow matching features and can resist vibration and thermal stresses during star-up and shutdown.

  8. Safe, Compact Nuclear Propulsion: Solid Core Nuclear Propulsion Concept

    Science.gov (United States)

    1988-10-01

    analysis group developed ROM component cost estimates given in representative ranges. 4.1 Engine System A representative nuclear thermal rocket engine...nuclear thermal rocket engine cycle balance computer code. The design requirements for the engine were: Thrust : 15,000 lbf Champer Pressure 500 psia...advanced nuclear thermal rockets . Our analysis was based on an examination of presentation material provided by Martin, some independent calculations of

  9. 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

  10. Open cycle gas core nuclear rockets

    Science.gov (United States)

    Ragsdale, Robert

    1991-01-01

    The open cycle gas core engine is a nuclear propulsion device. Propulsion is provided by hot hydrogen which is heated directly by thermal radiation from the nuclear fuel. Critical mass is sustained in the uranium plasma in the center. It has typically 30 to 50 kg of fuel. It is a thermal reactor in the sense that fissions are caused by absorption of thermal neutrons. The fast neutrons go out to an external moderator/reflector material and, by collision, slow down to thermal energy levels, and then come back in and cause fission. The hydrogen propellant is stored in a tank. The advantage of the concept is very high specific impulse because you can take the plasma to any temperature desired by increasing the fission level by withdrawing or turning control rods or control drums.

  11. Nuclear Human Resources Development Program using Educational Core Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Sun; Hong, Soon Kwan [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides.

  12. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Science.gov (United States)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  13. Development of Core Monitoring System for Nuclear Power Plants (I)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

  14. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  15. MOX fuel arrangement for nuclear core

    Science.gov (United States)

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  16. Modeling and design of a new core-moderator assembly and neutron beam ports for the Penn State Breazeale Nuclear Reactor (PSBR)

    Science.gov (United States)

    Ucar, Dundar

    This study is for modeling and designing a new reactor core-moderator assembly and new neutron beam ports that aimed to expand utilization of a new beam hall of the Penn State Breazeale Reactor (PSBR). The PSBR is a part of the Radiation Science and Engineering Facility (RSEC) and is a TRIGA MARK III type research reactor with a movable core placed in a large pool and is capable to produce 1MW output. This reactor is a pool-type reactor with pulsing capability up to 2000 MW for 10-20 msec. There are seven beam ports currently installed to the reactor. The PSBR's existing core design limits the experimental capability of the facility, as only two of the seven available neutron beam ports are usable. The finalized design features an optimized result in light of the data obtained from neutronic and thermal-hydraulics analyses as well as geometrical constraints. A new core-moderator assembly was introduced to overcome the limitations of the existing PSBR design, specifically maximizing number of available neutron beam ports and mitigating the hydrogen gamma contamination of the neutron beam channeled in the beam ports. A crescent-shaped moderator is favored in the new PSBR design since it enables simultaneous use of five new neutron beam ports in the facility. Furthermore, the crescent shape sanctions a coupling of the core and moderator, which reduces the hydrogen gamma contamination significantly in the new beam ports. A coupled MURE and MCNP5 code optimization analysis was performed to calculate the optimum design parameters for the new PSBR. Thermal-hydraulics analysis of the new design was achieved using ANSYS Fluent CFD code. In the current form, the PSBR is cooled by natural convection of the pool water. The driving force for the natural circulation of the fluid is the heat generation within the fuel rods. The convective heat data was generated at the reactor's different operating powers by using TRIGSIMS, the fuel management code of the PSBR core. In the CFD

  17. Design of an organic simplified nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

    2016-08-15

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  18. Design of an Organic Simplified Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  19. 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.

  20. Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

    Science.gov (United States)

    Weinstein, H.; Lavan, Z.

    1975-01-01

    Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

  1. Development of Few Group Cross Section Calculation Model for Core Nuclear Design Code CYCAS%堆芯核设计程序CYCAS少群截面模型开发

    Institute of Scientific and Technical Information of China (English)

    杨伟焱; 汤春桃; 毕光文; 杨波

    2016-01-01

    少群截面模型为堆芯三维扩散计算提供实时的节块均匀少群截面,是堆芯计算程序的关键模型之一.CYCAS程序是上海核工程研究设计院最新开发的堆芯三维核设计程序.本文在详细解析影响节块截面的各种因素的基础上,提出应用于CYCAS程序的少群截面的模型.该模型采用能谱修正方法处理由于能谱变化所引入的二次效应,采用微观燃耗修正方法处理燃耗历史效应.单组件和A P1000核电厂的数值验证计算表明,该模型具有很高的计算精度.%The few group cross section calculation model generates node homogeneous few group cross section for core 3D diffusion calculation ,w hich is one of the key models of core calculation code .CYCAS is the new core 3D nuclear design code developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI) .A new model based on detail analysis of the factors affecting node cross section was developed for CYCAS .In the model ,the energy spectrum correction method was used to process the second order effect introduced by energy spectrum change , and the micro-depletion correction method was utilized to treat depletion history effect .The numerical results of unit assembly and AP1000 core validate the high accuracy of the new model within CYCAS .

  2. Core design and optimization of high performance low sodium void 1000 MWe heterogeneous oxide LMFBR cores

    Energy Technology Data Exchange (ETDEWEB)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Beitel, J.C.; Turski, R.; Lam, P.S.K.; Fuller, E.L.

    1979-01-01

    Radially heterogeneous core configurations are effective means to reduce sodium void reactivity. In general, radially heterogeneous cores can be designed as tightly or loosely coupled cores with center core or center blanket arrangements. Core height, number of core regions and number of fuel pins per assembly are additional variables in an optimization of basic heterogeneous core configurations. An extensive study was carried out to optimize the core configurations for 1000 MWe LMFBRs. All cores were subject to a common set of nuclear, mechanical, and thermal-hydraulic design assumptions. They were restrained by an upper sodium void reactivity limit of $2.50 and a doubling time of approximately 15 to 18 years. The screening and optimization procedures employed lead to two core layouts which were both tightly coupled. A complete nuclear analysis of these two cores (derived from a loosely coupled configuration/derived from a tightly coupled configuration) determined the fissile inventories (4268.4/4213.4 kg at BOEC), burnups (83.90/100.7 MWd/t peak), reactivity swings (0.49/1.8% ..delta..k total), power and flux distributions for different control insertion patterns, the breeding performance (15.7/15.3 yrs CSDT), the safety parameters, such as sodium void reactivity ($2.38/$2.23 at EOEC), isothermal Doppler coefficients for both sodium-in (45.6/46.1 T dk/dT x 10/sup -4/ core at EOEC) and sodium-out conditions (28.6/28.2 T dk/dT x 10/sup -4/ core at EOEC), and the transient behavior which shows very little space-dependence during a 60 cent reactivity step insertion.

  3. A New Capability for Nuclear Thermal Propulsion Design

    Science.gov (United States)

    Amiri, Benjamin W.; Kapernick, Richard J.; Sims, Bryan T.; Simpson, Steven P.

    2007-01-01

    This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. The core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.

  4. The APR1400 Core Design by Using APA Code System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Sun [Korea Electric Power Research Institue, Daejeon (Korea, Republic of); Koh, Byung Marn [USERS, Daejeon (Korea, Republic of)

    2008-05-15

    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.

  5. 核电厂反应堆堆芯中子与温度探测器组件研制%Design of In-Core Neutron and Temperature Detector Assembly for Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    黄国良

    2014-01-01

    针对ACP1000堆型,研制了用于反应堆堆芯核测系统的堆芯中子和温度测量探测器组件。文介绍了探测器组件的设计、性能指标和试验结果。设计的堆芯中子和温度探测器组件集成了中子自给能探测器和测温元件并固定安装在堆内。试验结果表明测量敏感元件的性能满足设计要求,外壳和密封组件能保证反应堆一回路压力边界的要求。堆芯测量探测器组件一体化的设计可提高安全性和可靠性,实现实时测量,可用于反应堆保护。%For the ACP1000 reactor type , the detector assembly of measuring in -core neutron and temperature for reactor in-core nuclear detection system is designed .This article introduces the design , capability and ex-perimental results of this detector assembly .Detector assembly for measuring in -core neutron and temperature integrates self -powered neutron detectors and temperature measuring components and install them inside the reactor .Results indicate that the capability of sensitive components meets the design requirements , and shell and sealing assembly can meet the requirements of reactor loop pressure boundary .Integration of in -core measuring systems can improve the security and reliability , can achieve real -time measurements and can be used for reactor protection .

  6. Nuclear statistical equilibrium at core-collapse supernova

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    A new improved nuclear partition function is employed to calculate the nuclear statistical equilibrium (NSE) in core-collapse supernova environment. The results show that the change of nucleus abundance is slight even though the temperature is higher than 1011 K when shock propagates, which indicates that the effect of the nuclear partition function is not so important as shown in the previous calculations, but it can also be considered in detailed simulation if it is sensitive to weak interaction rates in core-collapse supernova.

  7. 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)

  8. 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.

  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...... through the use of micro-benchmarks that our principles guide the design of a processor core that improves performance by an average of 38% over a similar Xilinx MicroBlaze configuration....

  10. Reducing the risk to Mars: The gas core nuclear rocket

    Science.gov (United States)

    Howe, S. D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1998-01-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas-core nuclear rocket (GCNR) has the potential to be such a system. The authors have completed a comparative study of the potential impact that a GCNR could have on a manned Mars mission. The total IMLEO, transit times, and accumulated radiation dose to the crew will be compared with the NASA Design Reference Missions.

  11. Reactor core design and characteristics of the Fugen

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Mitsuo; Kowata, Yasuki; Sugawara, Satoru; Deshimaru, Takehide

    1988-03-01

    The heavy water moderated, boiling light water cooled pressure tube type reactor Fugen uses plutonium-uranium mixed oxide as a fuel. Heavy water as the moderator and the light water of coolant are separated by the pressure tubes and calandria tubes. Thereby, the reactor core is heterogenes compared with that of LWRs. This paper describes the development of reactor core design procedure based on the feature of the Fugen type reactor, the feasibility test and the validity of nuclear and thermalhydraulic design based on the operating experience.

  12. Gas core nuclear thermal rocket engine research and development in the former USSR

    Energy Technology Data Exchange (ETDEWEB)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G. [eds.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept.

  13. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

  14. 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...... on FPGA-based processor cores: first, superpipelining enables higher-frequency system clocks, and second, predicated instructions circumvent costly pipeline stalls due to branches. To evaluate their effects, we develop Tinuso, a processor architecture optimized for FPGA implementation. We demonstrate...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. The open-cycle gas-core nuclear rocket engine - Some engineering considerations.

    Science.gov (United States)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyk, L. C.

    1971-01-01

    A preliminary design study of a conceptual 6000-MW open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 44,200 lb and a specific impulse of 4400 sec. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel) and the waste heat rejection system were considered conceptually and were sized.

  17. Recent Problems of Transformer Core Design

    Science.gov (United States)

    Valkovic, Z.

    1988-01-01

    The paper describes the result of the investigations of the efficiency of power loss reduction in transformer cores made with high-permeability (HGO) and laser scribed (LS) grain-oriented electrical steels, and also the phenomena in three-limb three-phase cores with the so-called staggered T-joint design. The efficiency of the HGO material depends on core form and core induction. The efficiency is better for single-phase than for three-phase cores and also for higher induction. The localised efficiency of HGO material is not uniform and it is significantly lower in the yoke than in other parts. The efficiency of LS material (grade ZDKH) is better than that of the HGO material and also somewhat higher for single-phase than for three-phase cores. The localised flux distribution in the central limb of the core with staggered T-joint is more uniform and the content of higher harmonics is smaller than in the core with conventional V-45° T-joint. This results in a 13% loss reduction in the central limb and in a 4-5% reduction of total core loss.

  18. Robustness of nuclear core activity reconstruction by data assimilation

    Energy Technology Data Exchange (ETDEWEB)

    Bouriquet, Bertrand, E-mail: bertrand.bouriquet@cerfacs.f [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Argaud, Jean-Philippe [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Erhard, Patrick [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Massart, Sebastien [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Poncot, Angelique [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Ricci, Sophie [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Thual, Olivier [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Universite de Toulouse, INPT, UPS, IMFT, Allee Camille Soula, F-31400 Toulouse (France)

    2011-02-11

    We apply a data assimilation technique, inspired from meteorological applications, to perform an optimal reconstruction of the neutronic activity field in a nuclear core. Both measurements and information coming from a numerical model are used. We first study the robustness of the method when the amount of measured information decreases. We then study the influence of the nature of the instruments and their spatial repartition on the efficiency of the field reconstruction.

  19. Engineering design feasibility of low boron concentration core in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, A. T.; Kim, M. H. [Kyung Hee University, Yongin-shi, Gyeonggi-do, 446-701 Republic of Korea (Korea, Republic of); Woo, I.; Shon, S. R., E-mail: atdaing@khu.ac.k [Korea Nuclear Fuel, 1047 Daedukdaero, Yuseong-gu, Daejeon, 305-353 Republic of Korea (Korea, Republic of)

    2010-10-15

    In pressurized water reactor operation, higher level of soluble boron concentration could contribute higher impact from boron dilution situations, higher amount of liquid waste, and higher radiation dose to operators from higher corrosion potential to cladding and structure. Two practical and feasible means to reduce the maximum boron concentration were investigated in this study. A technically straightforward, possible means, can be achieved either by implementation of enriched boric acid (Eba) or by increasing more shim rod (fixed burnable absorber) worth. A simplest option is that the Eba is applied into reference core (Ref) design, OPR-1000 design, Ulchin unit-5 by allowing use of same fuel assemblies and core design without changing any nuclear design methodology used in that Ref design. Although results of Eba option proved its favorable power distribution and peaking factor, its moderator temperature coefficient (MTC) value reached positive, 3.25 pcm/ C at 40 EFPD which is beyond the design safety limit. An alternative option with more shim rods in fuel assemblies was tried with four types of integral burnable absorbers: gadolinia, integral fuel burnable absorber (Ifba), erbium and alumina boron carbide. Four core design candidates have been developed by keeping major engineering designs and preserving equivalent fuel enrichment level used in Ref design. However, all optimal designs were targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between Ref and optimal core designs is presented here. One of them is suggested as the most promising and favorable low boron core (Lbc) design in this framework. The proper combination of axial and radial enrichment zoning pattern in Lbc design candidate with Ifba-bearing fuel assemblies at equilibrium cycle, could bring 2 times narrower axial offset variation than that of Ref design, and maintain acceptable power peaking factor around 23% lower than

  20. Analysis of suprathermal nuclear processes in the solar core plasma

    Science.gov (United States)

    Voronchev, Victor T.; Nakao, Yasuyuki; Watanabe, Yukinobu

    2017-04-01

    A consistent model for the description of suprathermal processes in the solar core plasma naturally triggered by fast particles generated in exoergic nuclear reactions is formulated. This model, based on the formalism of in-flight reaction probability, operates with different methods of treating particle slow-down in the plasma, and allows for the influence of electron degeneracy and electron screening on processes in the matter. The model is applied to examine slowing-down of 8.7 MeV α-particles produced in the {}7{Li}(p,α )α reaction of the pp chain, and to analyze suprathermal processes in the solar CNO cycle induced by them. Particular attention is paid to the suprathermal {}14{{N}}{(α ,{{p}})}17{{O}} reaction unappreciated in standard solar model simulations. It is found that an appreciable non-standard (α ,p) nuclear flow due to this reaction appears in the matter and modifies running of the CNO cycle in ∼95% of the solar core region. In this region at R> 0.1{R}ȯ , normal branching of nuclear flow {}14{{N}}≤ftarrow {}17{{O}}\\to {(}18{{F}})\\to {}18{{O}} transforms to abnormal sequential flow {}14{{N}}\\to {}17{{O}}\\to {(}18{{F}})\\to {}18{{O}}, altering some element abundances. In particular, nuclear network calculations reveal that in the outer core the abundances of 17O and 18O isotopes can increase by a factor of 20 as compared with standard estimates. A conjecture is made that other CNO suprathermal (α ,p) reactions may also affect abundances of CNO elements, including those generating solar neutrinos.

  1. 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

  2. Nuclear data for designing the IFMIF accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    The objective of the International Fusion Materials Irradiation Facility (IFMIF) and the design concept of the IFMIF accelerator system are described. The status of the nuclear data, especially for the deuteron-induced reactions, to qualify the system design is reviewed. The requests for the nuclear data compilation and/or evaluation are summarized. (author)

  3. Piezoelectric material for use in a nuclear reactor core

    Science.gov (United States)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

  4. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  5. Review of coaxial flow gas core nuclear rocket fluid mechanics

    Science.gov (United States)

    Weinstein, H.

    1976-01-01

    Almost all of the fluid mechanics research associated with the coaxial flow gas core reactor ended abruptly with the interruption of NASA's space nuclear program because of policy and budgetary considerations in 1973. An overview of program accomplishments is presented through a review of the experiments conducted and the analyses performed. Areas are indicated where additional research is required for a fuller understanding of cavity flow and of the factors which influence cold and hot flow containment. A bibliography is included with graphic material.

  6. Promising design options for the encapsulated nuclear heat source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Conway, L.; Carelli, M.D.; Dzodzo, M. [Westinghouse Science and Technology, Pittsburgh, PA (United States); Hossain, Q.; Brown, N.W. [Lawrence Livermore National Lab., CA (United States); Wade, D.C.; Sienick, J.J. [Argonne National Lab., IL (United States); Greenspan, E.; Kastenberg, W.E.; Saphier, D. [University of California Dept of Nuclear Engineering, Berkeley, CA (United States)

    2001-07-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  7. Nuclear equation of state for core-collapse supernova simulations with realistic nuclear forces

    Energy Technology Data Exchange (ETDEWEB)

    Togashi, H., E-mail: hajime.togashi@riken.jp [Nishina Center for Accelerator-Based Science, Institute of Physical and Chemical Research (RIKEN), 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Nakazato, K. [Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Takehara, Y.; Yamamuro, S.; Suzuki, H. [Department of Physics, Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan); Takano, M. [Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Department of Pure and Applied Physics, Graduate School of Advanced Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan)

    2017-05-15

    A new table of the nuclear equation of state (EOS) based on realistic nuclear potentials is constructed for core-collapse supernova numerical simulations. Adopting the EOS of uniform nuclear matter constructed by two of the present authors with the cluster variational method starting from the Argonne v18 and Urbana IX nuclear potentials, the Thomas–Fermi calculation is performed to obtain the minimized free energy of a Wigner–Seitz cell in non-uniform nuclear matter. As a preparation for the Thomas–Fermi calculation, the EOS of uniform nuclear matter is modified so as to remove the effects of deuteron cluster formation in uniform matter at low densities. Mixing of alpha particles is also taken into account following the procedure used by Shen et al. (1998, 2011). The critical densities with respect to the phase transition from non-uniform to uniform phase with the present EOS are slightly higher than those with the Shen EOS at small proton fractions. The critical temperature with respect to the liquid–gas phase transition decreases with the proton fraction in a more gradual manner than in the Shen EOS. Furthermore, the mass and proton numbers of nuclides appearing in non-uniform nuclear matter with small proton fractions are larger than those of the Shen EOS. These results are consequences of the fact that the density derivative coefficient of the symmetry energy of our EOS is smaller than that of the Shen EOS.

  8. Nuclear factor Y regulates ancient budgerigar hepadnavirus core promoter activity.

    Science.gov (United States)

    Shen, Zhongliang; Liu, Yanfeng; Luo, Mengjun; Wang, Wei; Liu, Jing; Liu, Wei; Pan, Shaokun; Xie, Youhua

    2016-09-16

    Endogenous viral elements (EVE) in animal genomes are the fossil records of ancient viruses and provide invaluable information on the origin and evolution of extant viruses. Extant hepadnaviruses include avihepadnaviruses of birds and orthohepadnaviruses of mammals. The core promoter (Cp) of hepadnaviruses is vital for viral gene expression and replication. We previously identified in the budgerigar genome two EVEs that contain the full-length genome of an ancient budgerigar hepadnavirus (eBHBV1 and eBHBV2). Here, we found eBHBV1 Cp and eBHBV2 Cp were active in several human and chicken cell lines. A region from nt -85 to -11 in eBHBV1 Cp was critical for the promoter activity. Bioinformatic analysis revealed a putative binding site of nuclear factor Y (NF-Y), a ubiquitous transcription factor, at nt -64 to -50 in eBHBV1 Cp. The NF-Y core binding site (ATTGG, nt -58 to -54) was essential for eBHBV1 Cp activity. The same results were obtained with eBHBV2 Cp and duck hepatitis B virus Cp. The subunit A of NF-Y (NF-YA) was recruited via the NF-Y core binding site to eBHBV1 Cp and upregulated the promoter activity. Finally, the NF-Y core binding site is conserved in the Cps of all the extant avihepadnaviruses but not of orthohepadnaviruses. Interestingly, a putative and functionally important NF-Y core binding site is located at nt -21 to -17 in the Cp of human hepatitis B virus. In conclusion, our findings have pinpointed an evolutionary conserved and functionally critical NF-Y binding element in the Cps of avihepadnaviruses.

  9. Design of an Organic Simplified Nuclear Reactor

    OpenAIRE

    Koroush Shirvan; Eric Forrest

    2016-01-01

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attr...

  10. Designing the Nuclear Energy Attitude Scale.

    Science.gov (United States)

    Calhoun, Lawrence; And Others

    1988-01-01

    Presents a refined method for designing a valid and reliable Likert-type scale to test attitudes toward the generation of electricity from nuclear energy. Discusses various tests of validity that were used on the nuclear energy scale. Reports results of administration and concludes that the test is both reliable and valid. (CW)

  11. Designing the Nuclear Energy Attitude Scale.

    Science.gov (United States)

    Calhoun, Lawrence; And Others

    1988-01-01

    Presents a refined method for designing a valid and reliable Likert-type scale to test attitudes toward the generation of electricity from nuclear energy. Discusses various tests of validity that were used on the nuclear energy scale. Reports results of administration and concludes that the test is both reliable and valid. (CW)

  12. Validation of Fuqing Nuclear Power Plant Unit 1 Cycle 2 Refueling Design

    Institute of Scientific and Technical Information of China (English)

    PAN; Cui-jie; XIA; Zhao-dong; ZHU; Qing-fu

    2015-01-01

    Fuqing Nuclear Power Plant Unit 1Cycle 2refueling design was validated with the PWR core fuel management package CMS(CASMO5,CMSLINK5and SIMULATE5),including validating fuel management report,validating reload safety evaluation report,validating nuclear design report and validating physics tests report.

  13. Design data and safety features of commerical nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Heddleson, F.A.

    1976-06-01

    Design data, safety features, and site characteristics are summarized for 34 nuclear power units in 17 power stations in the United States. Six pages of data are presented for each plant, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. This volume covers Light Water Reactors (LWRs) with dockets 50-508 through 50-549, four HTGRs--50-171, 50-267, 50-450/451, 50-463/464, the Atlantic Floating Station 50-477/478, and the Clinch River Breeder 50-537.

  14. Virtual environments for nuclear power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W. [and others

    1996-03-01

    In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP).

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  17. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  18. HTGR nuclear heat source component design and experience

    Energy Technology Data Exchange (ETDEWEB)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included.

  19. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  20. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Hultqvist, Pontus

    2004-09-01

    In this thesis the possibility of locating vibrating noise sources in a nuclear reactor core from the neutron noise has been investigated using different localization methods. The influence of the vibrating noise source has been considered to be a small perturbation of the neutron flux inside the reactor. Linear perturbation theory has been used to construct the theoretical framework upon which the localization methods are based. Two different cases have been considered: one where a one-dimensional one-group model has been used and another where a two-dimensional two-energy group noise simulator has been used. In the first case only one localization method is able to determine the position with good accuracy. This localization method is based on finding roots of an equation and is sensitive to other perturbations of the neutron flux. It will therefore work better with the assistance of approximative methods that reconstruct the noise source to determine if the results are reliable or not. In the two-dimensional case the results are more promising. There are several different localization techniques that reproduce both the vibrating noise source position and the direction of vibration with enough precision. The approximate methods that reconstruct the noise source are substantially better and are able to support the root finding method in a more constructive way. By combining the methods, the results will be more reliable.

  1. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Directory of Open Access Journals (Sweden)

    Andrea Cerutti

    Full Text Available Hepatitis C virus (HCV infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS, but no nuclear export signal (NES has yet been identified.We show here that the aa(109-133 region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126 in the identified NES or in the sequence encoding the mature core aa(1-173 significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  2. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Science.gov (United States)

    Cerutti, Andrea; Maillard, Patrick; Minisini, Rosalba; Vidalain, Pierre-Olivier; Roohvand, Farzin; Pecheur, Eve-Isabelle; Pirisi, Mario; Budkowska, Agata

    2011-01-01

    Hepatitis C virus (HCV) infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS), but no nuclear export signal (NES) has yet been identified.We show here that the aa(109-133) region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126) in the identified NES or in the sequence encoding the mature core aa(1-173) significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  3. DIRSIG 5: core design and implementation

    Science.gov (United States)

    Goodenough, Adam A.; Brown, Scott D.

    2012-06-01

    The Digital Imaging and Remote Sensing Image Generation (DIRSIG) model has been developed at the Rochester Institute of Technology (RIT) for over two decades. The last major update of the model, DIRSIG 4, built on an established, first-principles, multi- and hyper-spectral scene simulation tool. It introduced a modern and flexible software architecture to support new sensor modalities and more complex and dynamic scenes. Since that time, the needs of the user community have grown and diversified in tandem with the computational capabilities of modern hardware. Faced with a desire to model more complex, multi-component systems that are beyond the original intent and capabilities of an aging software design, a new version of DIRSIG, version 5, is being introduced to the community. This paper describes the core of DIRSIG 5 that is responsible for linking the disparate sensor, scene, and environmental models together, spatially, temporally, and parametrically. The spatial relationships are governed by a planet-centric universe model encompassing a whole globe digital elevation and optical property model, the scene model(s), globally varying atmospheric models, and a space model. Temporal relationships are driven by a formal modeling and simulation architecture based on approaches used in engineering and biological sciences to model highly dynamic and interactive systems. Finally, the parametric interfaces are described by a universal data model that facilitates scripting, inter-dependent properties and user interface construction. The design of these components will be presented along with specific module implementation details. These simulation tools will be used to demonstrate some of the new capabilities and applications of DIRSIG 5.

  4. Nuclear-Powered GPS Spacecraft Design Study

    Energy Technology Data Exchange (ETDEWEB)

    Raab, Bernard

    1977-05-01

    This is the final report of a study to investigate the potential benefits of a nuclear (radioisotope) - powered satellite for advanced phases of the Global Positioning System (GPS) program. The critical parameters were: power to user; mean mission duration; orbital predictability; thermal control of on-board frequency standards; and vulnerability. The reference design approach is described, and input data are given for two power systems that are under development: an organic Rankine system and a Brayton cycle system. Reference design details are provided and structural design and analysis are discussed, as well as thermal design and analysis. A higher altitude version is also considered.

  5. Trends in nuclear systems design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G. (Commission of the European Communities, Joint Research Centre/Ispra Establishment, Inst. for Systems Engineering and Design (Italy))

    1991-12-01

    The papers presented at the poster session 'Nuclear Systems Design and Analyses' are reviewed and trends discussed. The attention is focussed on the following areas: (i) Reactor Studies and (ii) Neutronic Tests and Analysis. Present studies on D-T power reactor conceptual designs should be encouraged. Recent conceptual studies on D-{sup 3}He systems seem adequate to identify the critical issues. The economic feasibility of {sup 3}He supply from the moon should be assessed. Contributions on Neutronics are found to be of particular relevance and represent an important step towards the implementation of data base for nuclear design of plasma facing and breeding blanket components. (orig.).

  6. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  7. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    Science.gov (United States)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  8. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  9. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Energy Technology Data Exchange (ETDEWEB)

    Ansarifar, Gholam Reza; Saadatzi, Saeed [Dept. of Nuclear Engineering, Faculty of Advanced Sciences and Technology, University of Isfahan, Isfahan (Iran, Islamic Republic of)

    2015-12-15

    One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  10. Multiphysics Computational Analysis of a Solid-Core Nuclear Thermal Engine Thrust Chamber

    Science.gov (United States)

    Wang, Ten-See; Canabal, Francisco; Cheng, Gary; Chen, Yen-Sen

    2007-01-01

    The objective of this effort is to develop an efficient and accurate computational heat transfer methodology to predict thermal, fluid, and hydrogen environments for a hypothetical solid-core, nuclear thermal engine - the Small Engine. In addition, the effects of power profile and hydrogen conversion on heat transfer efficiency and thrust performance were also investigated. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics platform, while formulations of conjugate heat transfer were implemented to describe the heat transfer from solid to hydrogen inside the solid-core reactor. The computational domain covers the entire thrust chamber so that the afore-mentioned heat transfer effects impact the thrust performance directly. The result shows that the computed core-exit gas temperature, specific impulse, and core pressure drop agree well with those of design data for the Small Engine. Finite-rate chemistry is very important in predicting the proper energy balance as naturally occurring hydrogen decomposition is endothermic. Locally strong hydrogen conversion associated with centralized power profile gives poor heat transfer efficiency and lower thrust performance. On the other hand, uniform hydrogen conversion associated with a more uniform radial power profile achieves higher heat transfer efficiency, and higher thrust performance.

  11. Nuclear Safety Design Base for License Application

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2005-09-29

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111(b) [DIRS 173273] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113(b) and (c) [DIRS 173273] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period.

  12. Chemical Compound Design Using Nuclear Charge Distributions

    Science.gov (United States)

    2012-03-01

    NUMBER 6. AUTHOR(S) B. Christopher Rinderspacher 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING ORGANIZATION NAME(S...corresponds to the hydrogen atom. Hence, the set of non-negative distributions can be convexly decomposed and the associated projection onto nuclear...Xi, Y.; Saven, J. G. Advances in Computational Protein Design. Curr. Opin. Struct. Biol. 2004, 14 (4),487–494. [4] Mang , N. G.; Zeng, C. Reference

  13. Nuclear integrated database and design advancement system

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Jae Joo; Jeong, Kwang Sub; Kim, Seung Hwan; Choi, Sun Young

    1997-01-01

    The objective of NuIDEAS is to computerize design processes through an integrated database by eliminating the current work style of delivering hardcopy documents and drawings. The major research contents of NuIDEAS are the advancement of design processes by computerization, the establishment of design database and 3 dimensional visualization of design data. KSNP (Korea Standard Nuclear Power Plant) is the target of legacy database and 3 dimensional model, so that can be utilized in the next plant design. In the first year, the blueprint of NuIDEAS is proposed, and its prototype is developed by applying the rapidly revolutionizing computer technology. The major results of the first year research were to establish the architecture of the integrated database ensuring data consistency, and to build design database of reactor coolant system and heavy components. Also various softwares were developed to search, share and utilize the data through networks, and the detailed 3 dimensional CAD models of nuclear fuel and heavy components were constructed, and walk-through simulation using the models are developed. This report contains the major additions and modifications to the object oriented database and associated program, using methods and Javascript.. (author). 36 refs., 1 tab., 32 figs.

  14. Design considerations for an air core magnetic actuator

    Science.gov (United States)

    Groom, Nelson J.

    1992-01-01

    Equations for the force produced by an air core electromagnet on a permanent magnet core as a function of the coil height, coil inner and outer radii, and core displacement are developed. The magnetization vector of the permanent magnet core is assumed to be aligned with the central axis of the electromagnet and the forces which are produced lie along the same axis. Variations in force due to changes in electromagnet parameters and core displacement are investigated and parameter plots which should be useful for coil design are presented.

  15. Summary of Prometheus Radiation Shielding Nuclear Design Analysis

    Energy Technology Data Exchange (ETDEWEB)

    J. Stephens

    2006-01-13

    This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

  16. Nucleoporins as components of the nuclear pore complex core structure and Tpr as the architectural element of the nuclear basket.

    Science.gov (United States)

    Krull, Sandra; Thyberg, Johan; Björkroth, Birgitta; Rackwitz, Hans-Richard; Cordes, Volker C

    2004-09-01

    The vertebrate nuclear pore complex (NPC) is a macromolecular assembly of protein subcomplexes forming a structure of eightfold radial symmetry. The NPC core consists of globular subunits sandwiched between two coaxial ring-like structures of which the ring facing the nuclear interior is capped by a fibrous structure called the nuclear basket. By postembedding immunoelectron microscopy, we have mapped the positions of several human NPC proteins relative to the NPC core and its associated basket, including Nup93, Nup96, Nup98, Nup107, Nup153, Nup205, and the coiled coil-dominated 267-kDa protein Tpr. To further assess their contributions to NPC and basket architecture, the genes encoding Nup93, Nup96, Nup107, and Nup205 were posttranscriptionally silenced by RNA interference (RNAi) in HeLa cells, complementing recent RNAi experiments on Nup153 and Tpr. We show that Nup96 and Nup107 are core elements of the NPC proper that are essential for NPC assembly and docking of Nup153 and Tpr to the NPC. Nup93 and Nup205 are other NPC core elements that are important for long-term maintenance of NPCs but initially dispensable for the anchoring of Nup153 and Tpr. Immunogold-labeling for Nup98 also results in preferential labeling of NPC core regions, whereas Nup153 is shown to bind via its amino-terminal domain to the nuclear coaxial ring linking the NPC core structures and Tpr. The position of Tpr in turn is shown to coincide with that of the nuclear basket, with different Tpr protein domains corresponding to distinct basket segments. We propose a model in which Tpr constitutes the central architectural element that forms the scaffold of the nuclear basket.

  17. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

  18. Design, synthesis and applications of core-shell, hollow core, and nanorattle multifunctional nanostructures.

    Science.gov (United States)

    El-Toni, Ahmed Mohamed; Habila, Mohamed A; Labis, Joselito Puzon; ALOthman, Zeid A; Alhoshan, Mansour; Elzatahry, Ahmed A; Zhang, Fan

    2016-02-01

    With the evolution of nanoscience and nanotechnology, studies have been focused on manipulating nanoparticle properties through the control of their size, composition, and morphology. As nanomaterial research has progressed, the foremost focus has gradually shifted from synthesis, morphology control, and characterization of properties to the investigation of function and the utility of integrating these materials and chemical sciences with the physical, biological, and medical fields, which therefore necessitates the development of novel materials that are capable of performing multiple tasks and functions. The construction of multifunctional nanomaterials that integrate two or more functions into a single geometry has been achieved through the surface-coating technique, which created a new class of substances designated as core-shell nanoparticles. Core-shell materials have growing and expanding applications due to the multifunctionality that is achieved through the formation of multiple shells as well as the manipulation of core/shell materials. Moreover, core removal from core-shell-based structures offers excellent opportunities to construct multifunctional hollow core architectures that possess huge storage capacities, low densities, and tunable optical properties. Furthermore, the fabrication of nanomaterials that have the combined properties of a core-shell structure with that of a hollow one has resulted in the creation of a new and important class of substances, known as the rattle core-shell nanoparticles, or nanorattles. The design strategies of these new multifunctional nanostructures (core-shell, hollow core, and nanorattle) are discussed in the first part of this review. In the second part, different synthesis and fabrication approaches for multifunctional core-shell, hollow core-shell and rattle core-shell architectures are highlighted. Finally, in the last part of the article, the versatile and diverse applications of these nanoarchitectures in

  19. Design, synthesis and applications of core-shell, hollow core, and nanorattle multifunctional nanostructures

    Science.gov (United States)

    El-Toni, Ahmed Mohamed; Habila, Mohamed A.; Labis, Joselito Puzon; Alothman, Zeid A.; Alhoshan, Mansour; Elzatahry, Ahmed A.; Zhang, Fan

    2016-01-01

    With the evolution of nanoscience and nanotechnology, studies have been focused on manipulating nanoparticle properties through the control of their size, composition, and morphology. As nanomaterial research has progressed, the foremost focus has gradually shifted from synthesis, morphology control, and characterization of properties to the investigation of function and the utility of integrating these materials and chemical sciences with the physical, biological, and medical fields, which therefore necessitates the development of novel materials that are capable of performing multiple tasks and functions. The construction of multifunctional nanomaterials that integrate two or more functions into a single geometry has been achieved through the surface-coating technique, which created a new class of substances designated as core-shell nanoparticles. Core-shell materials have growing and expanding applications due to the multifunctionality that is achieved through the formation of multiple shells as well as the manipulation of core/shell materials. Moreover, core removal from core-shell-based structures offers excellent opportunities to construct multifunctional hollow core architectures that possess huge storage capacities, low densities, and tunable optical properties. Furthermore, the fabrication of nanomaterials that have the combined properties of a core-shell structure with that of a hollow one has resulted in the creation of a new and important class of substances, known as the rattle core-shell nanoparticles, or nanorattles. The design strategies of these new multifunctional nanostructures (core-shell, hollow core, and nanorattle) are discussed in the first part of this review. In the second part, different synthesis and fabrication approaches for multifunctional core-shell, hollow core-shell and rattle core-shell architectures are highlighted. Finally, in the last part of the article, the versatile and diverse applications of these nanoarchitectures in

  20. FBR core design with the composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, M.W.

    Although calculations are preliminary, overall feasibility of an FBR core design with the composite fuel assembly has been demonstrated. The advantaged over the heterogeneous design is that large variances in assembly mixed mean outlet temperatures are eliminated. Also, the effective enrichment of an assembly may easily be adjusted by varying the number of fertile pins per assembly, thus making it possible to flatten the core radial power profile. The use of the composite fuel assembly may in the future offer a significant alternative to heterogeneous FBR core design.

  1. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  2. Nuclear cardiology core syllabus of the European Association of Cardiovascular Imaging (EACVI).

    Science.gov (United States)

    Gimelli, Alessia; Neglia, Danilo; Schindler, Thomas H; Cosyns, Bernard; Lancellotti, Patrizio; Kitsiou, Anastasia

    2015-04-01

    The European Association of Cardiovascular Imaging (EACVI) Core Syllabus for Nuclear Cardiology is now available online. The syllabus lists key elements of knowledge in nuclear cardiology. It represents a framework for the development of training curricula and provides expected knowledge-based learning outcomes to the nuclear cardiology trainees. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2015. For permissions please email: journals.permissions@oup.com.

  3. Hanging core support system for a nuclear reactor. [LMFBR

    Science.gov (United States)

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

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  4. Simulation of Thermopower Influence on Fuel Core of Power Rod in Nuclear Power Plant (NPP Active Zone

    Directory of Open Access Journals (Sweden)

    I. S. Kulikov

    2010-01-01

    Full Text Available The paper considers problems of modern methods for  calculation of designs and materials of nuclear power. A model of numerical analysis for stress-strain state of fuel pins in the NPP active zone is proposed in the paper. The paper contains simulation concerning a fuel core section of a nuclear reactor heat-generating element with subsequent solution of a temperature and thermoelastic problem in computer program complex FEA ANSYS Workbench 11.0. All the obtained results have passed through checking procedure.

  5. Differential influence of instruments in nuclear core activity evaluation by data assimilation

    Energy Technology Data Exchange (ETDEWEB)

    Bouriquet, Bertrand, E-mail: bertrand.bouriquet@cerfacs.f [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Argaud, Jean-Philippe [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Erhard, Patrick [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Massart, Sebastien [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Poncot, Angelique [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Ricci, Sophie [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Thual, Olivier [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Universite de Toulouse, INPT, UPS, IMFT, Allee Camille Soula, F-31400 Toulouse (France)

    2011-01-21

    The global neutronic activity fields of a nuclear core can be reconstructed using data assimilation. Indeed, data assimilation allows to combine both measurements from instruments and information from a model, to evaluate the best possible neutronic activity within the core. We present and apply a specific procedure which evaluates the influence of measures by adding or removing instruments in a given measurement network (possibly empty). The study of various network configurations for the instruments in the nuclear core establishes that the influence of the instruments depends both on the independent instrumentation location and on the chosen network.

  6. Preliminary engineering design of sodium-cooled CANDLE core

    Science.gov (United States)

    Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

    2012-06-01

    The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CADLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

  7. 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...

  8. 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)

  9. Uranium droplet nuclear reactor core with MHD generator

    Science.gov (United States)

    Anghaie, Samim; Kumar, Ratan

    An innovative concept employing liquid uranium droplets as fuel in an ultrahigh-temperature vapor core reactor (UTVR) magnetohydrodynamic (MHD) generator power system for space power generation has been studied. Metallic vapor in superheated form acts as a working fluid for a closed-Rankine-type thermodynamic cycle. Usage of fuel and working fluid in this form assures certain advantages. The major technical issues emerging as a result involve a method for droplet generation, droplet transport in the reactor core, heat generation in the fuel and transport to the metallic vapor, and materials compatibility. A qualitative and quantitative attempt to resolve these issues has indicated the promise and tentative feasibility of the system.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  11. 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.

  12. Unified nuclear core activity map reconstruction using heterogeneous instruments with data assimilation

    CERN Document Server

    Bouriquet, Bertrand; Erhard, Patrick; Ponçot, Angélique

    2011-01-01

    Evaluating the neutronic state of the whole nuclear core is a very important topic that have strong implication for nuclear core management and for security monitoring. The core state is evaluated using measurements. Usually, part of the measurements are used, and only one kind of instruments are taken into account. However, the core state evaluation should be more accurate when more measurements are collected in the core. But using information from heterogeneous sources is at glance a difficult task. This difficulty can be overcome by Data Assimilation techniques. Such a method allows to combine in a coherent framework the information coming from model and the one coming from various type of observations. Beyond the inner advantage to use heterogeneous instruments, this leads to obtain a significant increasing of the quality of neutronic global state reconstruction with respect to individual use of measures. In order to present this approach, we will introduce here the basic principles of data assimilation f...

  13. Pinning down nuclear. To the core of the matter

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, Helmut; Gerstmayr, Michael [Technische Univ., Vienna (Austria); International Atomic Energy Agency, Vienna (Austria); Radde, Eileen [Nuclear Engineering Seibersdorf GmbH (Austria); International Atomic Energy Agency, Vienna (Austria)

    2014-07-01

    The nuclear disaster in Fukushima shocked the world tremendously. The call to pull out of nuclear energy is getting louder - and more often than not by politicians trying to lure the favour of voters. Through the media there are half-truths and false information floating about the global consequences of the disaster and sensational prognoses for the future, all of which are in turn unsettling for the general public. Are the opposers to nuclear energy playing with the fear of the public or is the threat real? This book tells, in a captivating manner - authenticated with examples and incidents not known by many - what the threat for the area actually looks like. They confront the level of truth in the frightening scenarios and inform about the situation in case of emergency. Furthermore, they examine factors that preceded the disaster and broach the subject of the incredible hunger for energy, which dominates the world and continues to drive the commercial use of nuclear energy. Also the ghost of Chernobyl and its aftermath, which has been dismissed from our minds, is re-examined based on current knowledge. The book impresses with insider know-how, latest detailed knowledge, amazing facts and an entertaining narrative style.

  14. Nuclear Fusion in the Deuterated cores of inflated hot Jupiters

    CERN Document Server

    Ouyed, Rachid

    2015-01-01

    In Ouyed et al. (1998), Deuterium-Deuterium (DD) burning in the deep interior of giant planets (at the core-mantle interface) was proposed as a mechanism to explain their observed heat excess. An issue with such a mechanism is the extreme condition of high interior temperatures (~ 10^5 K) in a concentrated D layer needed to account for the excess heat. In this paper, we show that screened DD fusion in a deuterated core is a more plausible mechanism to explain the excess heat and observed inflated radii of some Jovian exoplanets ("hot Jupiters"). The screening alleviates the extreme temperature constraint and removes the requirement of a stratified D layer, so that DD-fusion is a significant internal energy source (~ 10^(25)-10^(27) erg/s) even within the expected range of core temperature (~ 10^4 K) and density of hot Jupiters. The mechanism is universal, long-lasting (Gigayears), and should be effective as long as the metallicity is not too high and the core has not been significantly eroded away already. Ap...

  15. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  16. Russian Nuclear Rocket Engine Design for Mars Exploration

    Institute of Scientific and Technical Information of China (English)

    Vadim Zakirov; Vladimir Pavshook

    2007-01-01

    This paper is to promote investigation into the nuclear rocket engine (NRE) propulsion option that is considered as a key technology for manned Mars exploration. Russian NRE developed since the 1950 s in the former Soviet Union to a full-scale prototype by the 1990 s is viewed as advantageous and the most suitable starting point concept for manned Mars mission application study. The main features of Russian heterogeneous core NRE design are described and the most valuable experimental performance results are summarized. These results have demonstrated the significant specific impulse performance advantage of the NRE over conventional liquid rocket engine (LRE) propulsion technologies. Based on past experience,the recent developments in the field of high-temperature nuclear fuels, and the latest conceptual studies, the developed NRE concept is suggested to be upgraded to the nuclear power and propulsion system (NPPS),more suitable for future manned Mars missions. Although the NRE still needs development for space application, the problems are solvable with additional effort and funding.

  17. Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, Hak Sung; Park, Cheol [KAERI, Daejeon (Korea, Republic of); Nghiem, Huynh Ton; Vinh, Le Vinh; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    A conceptual nuclear design of a 20 MW multi-purpose research reactor for Vietnam has been jointly done by the KAERI and the DNRI (VAEC). The AHR reference core in this report is a right water cooled and a heavy water reflected open-tank-in-pool type multipurpose research reactor with 20 MW. The rod type fuel of a dispersed U{sub 3}Si{sub 2}-Al with a density of 4.0 gU/cc is used as a fuel. The core consists of fourteen 36-element assemblies, four 18-element assemblies and has three in-core irradiation sites. The reflector tank filled with heavy water surrounds the core and provides rooms for various irradiation holes. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worths, etc. For the analysis, the MCNP, MVP, and HELIOS codes were used by KAERI and DNRI (VAEC). The results by MCNP (KAERI) and MVP (DNRI) showed good agreements and can be summarized as followings. For a clean, unperturbed core condition such that the fuels are all fresh and there are no irradiation holes in the reflector region, the fast neutron flux (E{sub n}{>=}1.0 MeV) reaches 1.47x10{sup 14} n/cm{sup 2}s and the maximum thermal neutron flux (E{sub n}{<=}0.625 eV) reaches 4.43x10{sup 14} n/cm{sup 2}s in the core region. In the reflector region, the thermal neutron peak occurs about 28 cm far from the core center and the maximum thermal neutron flux is estimated to be 4.09x10{sup 14} n/cm{sup 2}s. For the analysis of the equilibrium cycle core, the irradiation facilities in the reflector region were considered. The cycle length was estimated as 38 days long with a refueling scheme of replacing three 36-element fuel assemblies or replacing two 36-element and one 18-element fuel assemblies. The excess reactivity at a BOC was 103.4 mk, and 24.6 mk at a minimum was reserved at an EOC. The assembly average discharge burnup was 54.6% of initial U-235 loading. For the proposed fuel management

  18. Multifunctional Core-Shell and Nano-channel Design for Nano-sized Thermo-sensor

    Science.gov (United States)

    2015-04-01

    L R E P O R T DTRA-TR-14-32 Multifunctional Core-Shell and Nano- channel Design for Nano-sized Thermo - sensor Distribution Statement A... Thermo -sensor PI: Jie Lian, Associate Professor, Department of Mechanical, Aerospace & Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY...within s time frame. (2) Scope This project is under the scope of Basic and Applied Sciences Directorate and the JSTO and Nano-sized Thermo -sensor

  19. 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

  20. (129)I record of nuclear activities in marine sediment core from Jiaozhou Bay in China.

    Science.gov (United States)

    Fan, Yukun; Hou, Xiaolin; Zhou, Weijian; Liu, Guangshan

    2016-04-01

    Iodine-129 has been used as a powerful tool for environmental tracing of human nuclear activities. In this work, a sediment core collected from Jiaozhou Bay, the east coast of China, in 2002 was analyzed for (129)I to investigate the influence of human nuclear activities in this region. Significantly enhanced (129)I level was observed in upper 70 cm of the sediment core, with peak values in the layer corresponding to 1957, 1964, 1974, 1986, and after 1990. The sources of (129)I and corresponding transport processes in this region are discussed, including nuclear weapons testing at the Pacific Proving Grounds, global fallout from a large numbers of nuclear weapon tests in 1963, the climax of Chinese nuclear weapons testing in the early 1970s, the Chernobyl accident in 1986, and long-distance dispersion of European reprocessing derived (129)I. The very well (129)I records of different human nuclear activities in the sediment core illustrate the potential application of (129)I in constraining ages and sedimentation rates of the recent sediment. The releases of (129)I from the European nuclear fuel reprocessing plants at La Hague (France) and Sellafield (UK) were found to dominate the inventory of (129)I in the Chinese sediments after 1990, not only the directly atmospheric releases of these reprocessing plants, but also re-emission of marine discharged (129)I of these reprocessing plants in the highly contaminated European seas.

  1. Nuclear design aspect of the Korean high intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa; Song, Tae-Yung [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

    1998-11-01

    A plan to construct a high current proton accelerator has been proposed by KAERI. We are presenting the required nuclear design to support the project as well as a brief overview of the proposed proton accelerator. The target and core design is highlighted to show feasibility of incineration of minor actinides from the spent fuel of light water reactors. Radiation shielding and activation analyses are also important for the design and the license of the accelerator. (author)

  2. Lunar mission design using Nuclear Thermal Rockets

    Science.gov (United States)

    Stancati, Michael L.; Collins, John T.; Borowski, Stanley K.

    1991-01-01

    The NERVA-class Nuclear Thermal Rocket (NTR), with performance nearly double that of advanced chemical engines, has long been considered an enabling technology for human missions to Mars. NTR engines address the demanding trip time and payload delivery needs of both cargo-only and piloted flights. But NTR can also reduce the Earth launch requirements for manned lunar missions. First use of NTR for the Moon would be less demanding and would provide a test-bed for early operations experience with this powerful technology. Study of application and design options indicates that NTR propulsion can be integrated with the Space Exploration Initiative scenarios to deliver performance gains while managing controlled, long-term disposal of spent reactors to highly stable orbits.

  3. Preliminary safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2001-06-01

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  4. Nuclear design report for Ulchin nuclear power plant unit 1, cycle 7

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Rae; Park, Yong soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-04-01

    This report presents nuclear design calculations for Cycle 7 of Ulchin Unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 56 KOFA`s enriched by nominally 4.00 w/o U{sub 235}. Among the KOFA`s 36 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 7 amounts to 355 EFPD corresponding to a cycle burnup of 14280 MWD/MTU. (Author) 8 refs., 55 figs., 21 tabs.

  5. Nuclear design report for Ulchin nuclear power plant unit 2 cycle 5

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ha; Park, Yong Soo; Cho, Byeong Ho; Zee, Sung Kyun; Lee, Sang Keun; Ahn, Dawk Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-09-01

    This report presents nuclear design calculations for cycle 5 of Ulchin unit it 2. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 48 KOFA`s enriched by nominally 3.50 w/o U{sub 235}. Among the KOFA`s, 20 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 5 amounts to 293 EFPD corresponding to a cycle burnup of 11780 MWD/MTU. (Author) 8 refs., 55 figs., 16 tabs.

  6. Nuclear design report for Yonggwang nuclear power plant unit 1 cycle 9

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young chul; Kim, Jae Hak; Song, Jae Woong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-03-01

    This report presents nuclear design calculations for Cycle 6 of Yonggwng Unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 4.00 w/o U{sub 235}. Among the KOFA`s, 60 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 9 amounts to 434 EFPD corresponding to a cycle burnup of 17470 MWD/MTU. (Author) 8 refs., 55 figs., 19 tabs.

  7. Nuclear design report for Kori nuclear power plant unit 4 cycle 8

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyoon; Jung, Yil Sub; Kim, Si Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-07-01

    This report presents nuclear design calculations for cycle 8 of Kori unit 4. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s 48 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 8 amounts to 421 EFPD corresponding to a cycle burnup of 16950 MWD/MTU. (Author) 8 refs., 55 figs., 17 tabs.

  8. Nuclear design report for Ulchin nuclear power plant unit 2, cycle 6

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chan Oh; Park, Jin Ha; Kim, Yong Rae; Park, Sang Yoon; Lee, Jong Chul; Baik, Joo Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    This report presents nuclear design calculations for cycle 6 of Ulchin unit 2. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 64 KOFA`s enriched by nominally 3.80 w/o U{sub 235}. Among the KOFA`s, 36 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 6 amounts to 388 EFPD corresponding to a cycle burnup of 15610 MWD/MTU. (Author) 8 refs., 55 figs., 17 tabs.

  9. Nuclear design report for Kori nuclear power plant unit 1, cycle 13

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Moon, Bok Ja; Cho, Byeong Ho; Jung, Yil Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-04-01

    This report presents nuclear design calculations for cycle 13 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 44 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 16 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 13 amounts to 355 EFPD corresponding to a cycle burnup of 13240 MWD/MTU. (Author) 8 refs., 55 figs., 16 tabs.

  10. Nuclear design report for Yonggwang nuclear power plant unit 1, cycle 8

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Chul; Kim, Jae Hak; Park, Sang Yoon; Zee, Sung Kyun; Lee, Sang Keun; Ahn, Dawk Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-10-01

    This report presents nuclear design calculations for cycle 8 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 56 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 8 amounts to 447 EFPD corresponding to a cycle burnup of 18020 MWD/MTU. (Author) 8 refs., 39 figs., 17 tabs.

  11. Nuclear design report for Ulchin nuclear power plant unit 1, cycle 6

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Kim, Yong Rae; Park, Yong Soo; Cho, Byeong Ho; Lee, Sang Keun; Ahn, Dawk Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    This report presents nuclear design calculations for cycle 6 of Ulchin unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 64 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 32 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 6 amounts to 369 EFPD corresponding to a cycle burnup of 14850 MWD/MTU. (Author) 8 refs., 55 figs., 17 tabs.

  12. Nuclear design report for Yonggwang nuclear power plant unit 1 cycle 9

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young chul; Kim, Jae Hak; Song, Jae Woong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-03-01

    This report presents nuclear design calculations for Cycle 6 of Yonggwng Unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 4.00 w/o U{sub 235}. Among the KOFA`s, 60 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 9 amounts to 434 EFPD corresponding to a cycle burnup of 17470 MWD/MTU. (Author) 8 refs., 55 figs., 19 tabs.

  13. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  14. Design of Broadband Single Fundamental Mode Hollow Core Bragg Fibre

    Institute of Scientific and Technical Information of China (English)

    LIN Chen-Xi; ZHANG Wei; HUANG Yi-Dong; PENG Jiang-De

    2008-01-01

    The condition of the single fundamental mode(HE11)transmission in hollow core Bragg fibres is investigated theoretically by the transfer matrix method.The influences of core size and cladding parameters on the single HE11 mode bandwidth are analysed,showing that the maximal bandwidth is more sensirive to the core size than the cladding.The numerical results show that sufficiently broad bandwidth of single HE11 mode transmission can be achieved by proper fibre design.A simple and fast method based on improved hollow metal waveguide model js proposed to optimize fibre structure parameters for the maximal single HE11 mode bandwidth.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  16. Mission design considerations for nuclear risk mitigation

    Science.gov (United States)

    Stancati, Mike; Collins, John

    1993-01-01

    Strategies for the mitigation of the nuclear risk associated with two specific mission operations are discussed. These operations are the safe return of nuclear thermal propulsion reactors to earth orbit and the disposal of lunar/Mars spacecraft reactors.

  17. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  18. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  19. Nuclear determination of saturation profiles in core plugs. Status report

    Energy Technology Data Exchange (ETDEWEB)

    Sletsgaard, J. [DTU, Inst. for Automation (Denmark)

    1996-01-01

    A method to determine liquid saturations in core plugs during flooding is of importance when the relative permeability and capillary pressure function are to be determined. This part of the EFP-93 project uses transmission of {gamma}-radiation to determine these saturations. In {gamma}-transmission measurements, the electron density of the given substance is measured. This is an advantage as compared to methods that use electric conductivity, since neither oil nor gas conducts electricity. At the moment a single {sup 137}Cs-source is used, but a theoretical investigation of whether it is possible to determine three saturations, using two radioactive sources with different {gamma}-energies, has been performed. Measurements were made on three core plugs. To make sure that the measurements could be reproduced, all the plugs had a point of reference, i.e. a mark so that it was possible to place the plug same way every time. Two computer programs for calculation of saturation and porosity and the experimental setup are listed. (EG).

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

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

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

  1. Analysis and Design of ITER 1 MV Core Snubber

    Institute of Scientific and Technical Information of China (English)

    王海田; 李格

    2012-01-01

    The core snubber, as a passive protection device, can suppress arc current and absorb stored energy in stray capacitance during the electrical breakdown in accelerating electrodes of ITER NBI. In order to design the core snubber of ITER, the control parameters of the arc peak current have been firstly analyzed by the Fink-Baker-Owren (FBO) method, which are used for designing the DIIID 100 kV snubber. The B-H curve can be derived from the measured voltage and current waveforms, and the hysteresis loss of the core snubber can be derived using the revised parallelogram method. The core snubber can be a simplified representation as an equivalent parallel resistance and inductance, which has been neglected by the FBO method. A simulation code including the parallel equivalent resistance and inductance has been set up. The simulation and experiments result in dramatically large arc shorting currents due to the parallel inductance effect. The case shows that the core snubber utilizing the FBO method gives more compact design.

  2. New Nuclear Equation of State for Core-Collapse Supernovae with the Variational Method

    Directory of Open Access Journals (Sweden)

    Togashi H.

    2014-03-01

    Full Text Available We report the current status of our project to construct a new nuclear equation of state (EOS with the variational method for core-collapse supernova (SN simulations. Starting from the realistic nuclear Hamiltonian, the EOS for uniform nuclear matter is constructed with the cluster variational method: For non-uniform nuclear matter, the EOS is calculated with the Thomas-Fermi method. The obtained thermodynamic quantities of uniform matter are in good agreement with those with more sophisticated Fermi Hypernetted Chain variational calculations, and phase diagrams constructed so far are close to those of the Shen-EOS. The structure of neutron stars calculated with this EOS at zero temperature is consistent with recent observational data, and the maximum mass of the neutron star is slightly larger than that with the Shen-EOS. Using the present EOS of uniform nuclear matter, we also perform the 1D simulation of the core-collapse supernovae by a simplified prescription of adiabatic hydrodynamics. The stellar core with the present EOS is more compact than that with the Shen-EOS, and correspondingly, the explosion energy in this simulation with the present EOS is larger than that with the Shen-EOS.

  3. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

    2008-09-15

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

  4. Dynamical Behavior of Core 3 He Nuclear Reaction-Diffusion Systems and Sun's Gravitational Field

    Institute of Scientific and Technical Information of China (English)

    DU Jiulin; SHEN Hong

    2005-01-01

    The coupling of the sun's gravitational field with processes of diffusion and convection exerts a significant influence on the dynamical behavior of the core 3He nuclear reaction-diffusion system. Stability analyses of the system are made in this paper by using the theory of nonequilibrium dynamics. It is showed that, in the nuclear reaction regions extending from the center to about 0.38 times of the radius of the sun, the gravitational field enables the core 3He nuclear reaction-diffusion system to become unstable and, after the instability, new states to appear in the system have characteristic of time oscillation. This may change the production rates of both 7Be and 8B neutrinos.

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

    Science.gov (United States)

    Clark, John S.; Walton, James T.; Mcguire, Melissa L.

    1992-01-01

    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.

  6. Highly Birefringent Photonic Crystal Fibers BUsing Asymmetric Core Design

    Institute of Scientific and Technical Information of China (English)

    Zhao Chun-Liu; Lu Chao; Yan Min; Wang Xiaoyan; Lou Junjun; Li Qin; Zhou Xiaoqun; Cai Qing; P.R.Chaudhuri

    2003-01-01

    We demonstrate a highly birefringent photonic crystal fiber by utilizing the asymmetric core design. Based on spectral measurements of the polarization mode interfering, we estimate that the fiber has a beat length of about 0.33 mm at 1545 nm.

  7. Planning of the development of the MMIS core technology based on nuclear-IT convergence

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Kim, Chang Hwoi; Hwang, In Koo [KAERI, Daejeon (Korea, Republic of); and others

    2012-01-15

    - Drive nuclear-IT convergence technologies such as middleware applied new concept nuclear instrumentation and control architecture, automated operation of future nuclear power plant, virtual reality/augmented reality, design and verification technology of a nuclear power plant main control room, software dependability, and cyber security technology - Write state-of-the-art report for the nuclear instrumentation and control based on IT convergence - A prototype which implemented related equipment and software subject to nuclear reactor operator that reside in the main control room (Reactor Operator, RO) order to a on-site operator (Local Operator, LO) and confirm the task performance matches the RO's intention - 'IT Convergence intelligent instrumentation and control technology' project planning for the Fourth Nuclear Power Research and Development in the long-term plan.

  8. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    2016-09-01

    The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

  9. Using a Genetic Algorithm to Design Nuclear Electric Spacecraft

    Science.gov (United States)

    Pannell, William P.

    2003-01-01

    The basic approach to to design nuclear electric spacecraft is to generate a group of candidate designs, see how "fit" the design are, and carry best design forward to the next generation. Some designs eliminated, some randomly modified and carried forward.

  10. Nuclear renaissance, public perception and design criteria: An exploratory review

    Energy Technology Data Exchange (ETDEWEB)

    Goodfellow, Martin J., E-mail: Martin.Goodfellow@postgrad.manchester.ac.uk [School of Chemical Engineering and Analytical Science, The Mill, Sackville Street, University of Manchester, Manchester M13 9PL (United Kingdom); Rolls-Royce, SINA-CNB-1, PO Box 31, Derby DE24 8BJ (United Kingdom); Williams, Hugo R. [Space Research Centre, University of Leicester, University Rd., Leicester LE1 7RH (United Kingdom); Azapagic, Adisa [School of Chemical Engineering and Analytical Science, The Mill, Sackville Street, University of Manchester, Manchester M13 9PL (United Kingdom)

    2011-10-15

    There is currently an international drive to build new nuclear power plants, bringing about what is being termed a 'nuclear renaissance'. However, the public perception of nuclear energy has historically been, and continues to be, a key issue, particularly in light of the Fukushima nuclear incident. This paper discusses the disparity between perceived and calculated risks based on the last four decades of research into risk perception. The leading psychological and sociological theories, Psychometric Paradigm and Cultural Theory, respectively, are critically reviewed. The authors then argue that a new nuclear-build policy that promotes a broader approach to design incorporating a wider range of stakeholder inputs, including that of the lay public, may provide a means for reducing the perceived risk of a nuclear plant. Further research towards such a new approach to design is proposed, based on integrating expert and lay stakeholder inputs and taking into account broader socio-cultural factors whilst maintaining the necessary emphasis on safety, technological development, economics and environmental sustainability. - Highlights: > Globally, a number of countries are investing in or considering building new nuclear plants. > Public acceptance of nuclear safety is important to continuing new nuclear build efforts. > Theories are discussed attempting to explain the public perception of nuclear safety. > A socially informed design process is proposed which could assist in ensuring public support. > Further research to understand how this design process might be performed is proposed.

  11. Accelerator-Driven Subcritical Fission in a Molten Salt Core: Green Nuclear Power for the New Millennium

    Science.gov (United States)

    McIntyre, Peter

    2011-10-01

    Scientists at Texas A&M University, Brookhaven National Lab, and Idaho National Lab are developing a design for accelerator-drive subcritical fission in a molten salt core (ADSMS). Three high-power proton beams are delivered to spallation targets in a molten salt core, where they provide ˜3% of the fast neutrons required to sustain 600 MW of fission. The proton beams are produced by a flux-coupled stack of superconducting strong-focusing cyclotrons. The fuel consists of a eutectic of sodium chloride with either spent nuclear fuel from a conventional U power reactor (ADSMS-U) or thorium (ADSMS-Th). The subcritical core cannot go critical under any failure mode. The core cannot melt down even if all power is suddenly lost to the facility for a prolonged period. The ultra-fast neutronics of the core makes it possible to operate in an isobreeding mode, in which neutron capture breeds the fertile nuclide into a fissile nuclide at the same rate that fission burns the fissile nuclide, and consumes 90% of the fertile inventory instead of the 5% consumed in the original use in a conventional power plant. The ultra-fast neutronics produces a very low equilibrium inventory of the long-lived minor actinides, ˜10^4 less than what is produced in conventional power plants. ADSMS offers a method to safely produce the energy needs for all mankind for the next 3000 years.

  12. SPACE-R Thermionic Space Nuclear Power System: Design and Technology Demonstration Program

    Science.gov (United States)

    1993-05-01

    This semiannual technical progress report summarizes the technical progress and accomplishments for the Thermionic Space Nuclear Power System (TI-SNPS) Design and Technology Demonstration Program of the prime contractor, Space Power Incorporated (SPI), its subcontractors, and supporting national laboratories during the first half of the government fiscal year (GFY) 1993. SPI's subcontractors and supporting national laboratories include: Babcock & Wilcox for the reactor core and externals; Space Systems/Loral for the spacecraft integration; Thermocore for the radiator heat pipes and the heat exchanger; INERTEK of CIS for the TFE, core elements, and nuclear tests; Argonne National Laboratories for nuclear safety, physics, and control verification; and Oak Ridge National laboratories for materials testing. Parametric trade studies are near completion. However, technical input from INERTEK has yet to be provided to determine some of the baseline design configurations. The INERTEK subcontract is expected to be initiated soon. The point design task has been initiated. The thermionic fuel element (TFE) is undergoing several design iterations. The reactor core vessel analysis and design has also been started.

  13. Optimization Design and Finite Element Analysis of Core Cutter

    Institute of Scientific and Technical Information of China (English)

    CAO Pin-lu; YIN Kun; PENG Jian-ming; LIU Jian-lin

    2007-01-01

    The hydro-hammer sampler is a new type of sampler compared with traditional ones. An important part of this new offshore sampler is that the structure of the core cutter has a significant effect on penetration and core recovery. In our experiments, a commercial finite element code with a capability of simulating large-strain frictional contact between two or more solid bodies is used to simulate the core cutter-soil interaction. The effects of the cutting edge shape, the diameter and the edge angle on penetration are analyzed by non-liner transient dynamic analysis using a finite element method (FEM). Simulation results show that the cutter shape clearly has an effect on the penetration and core recovery. In addition, the penetration of the sampler increases with an increase in the inside diameter of the cutter, but decreases with an increase in the cutting angle. Based on these analyses, an optimum structure of the core cutter is designed and tested in the north margin of the Dalian gulf. Experiment results show that the penetration rate is about 16.5 m/h in silty clay and 15.4 m/h in cohesive clay, while the recovery is 68% and 83.3% respectively.

  14. 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.

  15. 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)

  16. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  17. Chernobyl nuclear accident revealed from the 7010 m Muztagata ice core record

    Institute of Scientific and Technical Information of China (English)

    TIAN LiDe; YAO TanDong; WU GuangJian; LI Zhen; XU BaiQing; LI YueFang

    2007-01-01

    The total activity variation with depth from a 41.6 m Muztagata ice core drilled at 7010 m,recorded not only the 1963 radioactive layer due to the thermonuclear test,but also clearly the radioactive peak released by the Chernobyl accident in 1986.This finding indicates that the Chernobyl nuclear accident was clearly recorded in alpine glaciers in the Pamirs of west China,and the layer can be potentially used for ice core dating in other high alpine glaciers in the surrounding regions.

  18. Designed porosity materials in nuclear reactor components

    Science.gov (United States)

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  19. Designed porosity materials in nuclear reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  20. Design of the Core 2-4 GHz Betatron Equalizer

    Energy Technology Data Exchange (ETDEWEB)

    Deibele, C.; /Fermilab

    2000-01-01

    The core betatron equalizer in the Accumulator in the Antiproton Source at Fermilab needed to be upgraded. The performance could be rated as only circa 650 MHz when the system was a 2 GHz system. The old equalizer did not correct for the strong phase mismatch for the relatively strong gain of the system slightly below 2 GHz. The design corrects this phase mismatch and is relatively well matched both in and out of band.

  1. Comparative assessment of out-of-core nuclear thermionic power systems

    Science.gov (United States)

    Estabrook, W. C.; Koenig, D. R.; Prickett, W. Z.

    1975-01-01

    The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds.

  2. An assessment of coupling algorithms for nuclear reactor core physics simulations

    Science.gov (United States)

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss-Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton-Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  3. Design and construction of nuclear power plants

    CERN Document Server

    Schnell, Jürgen; Meiswinkel, Rüdiger; Bergmeister, Konrad; Fingerloos, Frank; Wörner, Johann-Dietrich

    2013-01-01

    Despite all the efforts being put into expanding renewable energy sources, large-scale power stations will be essential as part of a reliable energy supply strategy for a longer period. Given that they are low on CO2 emissions, many countries are moving into or expanding nuclear energy to cover their baseload supply.Building structures required for nuclear installations whose protective function means they are classified as safety-related, have to meet particular construction requirements more stringent than those involved in conventional construction. This book gives a comprehensive overv

  4. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Savrasov, Sergey [Univ. of California, Davis, CA (United States); Kotliar, Gabriel [Rutgers Univ., Piscataway, NJ (United States); Haule, Kristjan [Rutgers Univ., Piscataway, NJ (United States)

    2014-06-03

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  5. 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.

  6. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  7. A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible.

  8. Preliminary design report for SCDAP/RELAP5 lower core plate model

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Griffin, F.P. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a

  9. Neutrino-pair emission from nuclear de-excitation in core-collapse supernova simulations

    CERN Document Server

    Fischer, Tobias; Martinez-Pinedo, Gabriel

    2013-01-01

    We study the impact of neutrino-pair production from the de-excitation of highly excited heavy nuclei on core-collapse supernova simulations, following the evolution up to several 100 ms after core bounce. Our study is based on the AGILE-Boltztran supernova code, which features general relativistic radiation hydrodynamics and accurate three-flavor Boltzmann neutrino transport in spherical symmetry. In our simulations the nuclear de-excitation process is described in two different ways. At first we follow the approach proposed by Fuller and Meyer [Astrophys. J. 376,701 (1991)], which is based on strength functions derived in the framework of the nuclear Fermi-gas model of non-interacting nucleons. Secondly, we parametrize the allowed and forbidden strength distributions in accordance with measurements for selected nuclear ground states. We determine the de-excitation strength by applying the Brink hypothesis and detailed balance. For both approaches, we find that nuclear de-excitation has no effect on the supe...

  10. The Sensitivity of Core-Collapse Supernovae to Nuclear Electron Capture

    CERN Document Server

    Sullivan, Chris; Zegers, Remco G T; Grubb, Thomas; Austin, Sam M

    2015-01-01

    A weak-rate library aimed at investigating the sensitivity of astrophysical environments to variations of electron-capture rates on medium-heavy nuclei has been developed. With this library, the sensitivity of the core-collapse and early post-bounce phases of core-collapse supernovae to nuclear electron-capture is examined by systematically and statistically varying electron-capture rates of individual nuclei. The rates are adjusted by factors consistent with uncertainties indicated by comparing theoretical rates to those deduced from charge-exchange and $\\beta$-decay measurements. To ensure a model independent assessment, sensitivity studies across a comprehensive set of progenitors and equations of state are performed. In our systematic study, we find a +16/-4 % range in the mass of the inner-core at the time of shock formation and a $\\pm$20% range of peak {\

  11. KiloPower Project - KRUSTY Experiment Nuclear Design

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Godfroy, Thomas [NASA Marshall Space Flight Center (MSFC), Huntsville, AL (United States); Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-07-20

    This PowerPoint presentation covers the following topics: Reference Kilopower configuration; Reference KRUSTY configuration; KRUSTY design sensitivities; KRUSTY reactivity coefficients; KRUSTY criticality safety and control; KRUSTY core activation/dose; and KRUSTY shielding, room activation/dose.

  12. Nuclear design report for Kori nuclear power plant unit 1, cycle 14

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chan Oh; Kim, Joo Young; Park, Sang Yoon; Song, Jae Woong; Lee, Chong Chul; Baik, Joo Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report presents nuclear design calculations for cycle 14 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 44 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 16 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 14 amounts to 366 EFPD corresponding to a cycle burnup of 13680 MWD/MTU. (Author) 8 refs., 55 figs., 16 tabs. nozzle by vortex formation during mid-loop operation condition are experimentally investigated. The critical submergence is determined for various types of suction nozzle, and the measurements of velocity distribution are performed in the flow fields near the t-shaped suction nozzle. (Author) 11 refs., 41 figs., 13 tabs.

  13. The IPE Database: providing information on plant design, core damage frequency and containment performance

    Energy Technology Data Exchange (ETDEWEB)

    Lehner, J.R.; Lin, C.C.; Pratt, W.T. [Brookhaven National Lab., Upton, NY (United States); Su, T.; Danziger, L. [U.S. Nuclear Regulartory Commission, No. Bethesda, MD (United States)

    1996-08-01

    A database, called the IPE Database has been developed that stores data obtained from the Individual Plant Examinations (IPEs) which licensees of nuclear power plants have conducted in response to the Nuclear Regulatory Commission`s (NRC) Generic Letter GL88-20. The IPE Database is a collection of linked files which store information about plant design, core damage frequency (CDF), and containment performance in a uniform, structured way. The information contained in the various files is based on data contained in the IPE submittals. The information extracted from the submittals and entered into the IPE Database can be manipulated so that queries regarding individual or groups of plants can be answered using the IPE Database.

  14. Conceptual Design of the Top Mounted In-core Instrumentation for APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Taehyun; Bae, Jaehyun; Kim, Jongmin; Maeng, Cheolsoo; Kim, Hyunmin; Lee, Daehee [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

    2015-10-15

    Safety issue for nuclear power plant is reviewed. One of the main issue is Top Mounted In-core Instrumentation (TM-ICI). TM-ICI has the advantage of the structural integrity on the reactor bottom head during severe accident. This research about adopting the TM-ICI for APR1400 has been performed to have this advantage. Designing the nuclear power plant, safety issue is very important, and TM-ICI is one of the main issue. For the research of the TM-ICI, APR1400 has been reviewed for possibility of the TM-ICI and the results are as follows: The ICI nozzle head penetration shall be located outside of the CEDM nozzles and Two types of CEAs shall be unified into one type.

  15. Conceptual Design of a Nuclear Reactor Dedicated for Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yong Hun; Moon, Jang Sik; Jeong, Yong Hoon [Korea Adavanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The many advantages of nuclear desalination, the nuclear safety issues still remain a perennial problem today. To respond to such needs, the development of a desalination-dedicated nuclear reactor with maximized safety features was proposed. From the feasibility study, the desalination-dedicated reactor was found to be a good solution for meeting future water demand during the winter season in some countries like UAE by decoupling water and electricity supply. The economic analysis results indicated that under certain conditions, the desalination-dedicated reactor can produce freshwater at lower cost than the target nuclear cogeneration reactor using steam extraction technologies. A conceptual design of the desalination-dedicated nuclear reactor is in progress. The design features of the desalination-dedicated nuclear reactor could significantly enhance safety, reliability, and simplicity, and facilitate the extensive use of innovative passive safety systems. These maximized safety features of desalination-dedicated reactor could provide advanced capabilities for passive reactor shutdown and residual heat removal, and eventually prevent radioactivity release into the environment. The conceptual design achieved will provide a foothold for the future commercialization of the desalination-dedicated nuclear reactor and eventually help to address both a serious water crisis and nuclear safety issues.

  16. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    Science.gov (United States)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  17. Reactor Core Scheme for Small Nuclear Power Plant%小型核电站堆芯方案

    Institute of Scientific and Technical Information of China (English)

    解家春; 刘天才

    2012-01-01

    The small nuclear power planl enjoys advantages of long life and passive safely and is an important choice in the future nuclear power development. A conceptual core is designed for the small nuclear power planl. It is a pool-type fast reactor with sodium as coolant, the movable reflector and the fixed absorber as the reactivity control system for long-life. Further calculation results show thai the life of the reactor could be as long as 30 years, with a reasonable power distribution, all the reactivity coefficients negative, enough reactivity control worth, and all parameters satisfy the design requirements.%具有长寿命、非能动安全的小型核电站是核电发展的一个重要方向.本研究设计了一个小型核电站堆芯方案.该方案为池式钠冷快堆,采用移动反射层和堆内固定吸收体实现较长的堆芯寿期.进一步计算表明,该堆芯方案的寿期可达30年,功率分布合理,各种反应性系数为负值,控制方式的价值足够,满足设计要求.

  18. Human factor engineering applied to nuclear power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Manrique, A. [TECNATOM SA, BWR General Electric Business Manager, Madrid (Spain); Valdivia, J.C. [TECNATOM SA, Operation Engineering Project Manager, Madrid (Spain); Jimenez, A. [TECNATOM SA, Operation Engineering Div. Manager, Madrid (Spain)

    2001-07-01

    For the design and construction of new nuclear power plants as well as for maintenance and operation of the existing ones new man-machine interface designs and modifications are been produced. For these new designs Human Factor Engineering must be applied the same as for any other traditional engineering discipline. Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. Additionally, the big saving achieved by a nuclear power plant having an operating methodology which significantly decreases the risk of operating errors makes it necessary and almost vital its implementation. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. (author)

  19. Understanding seismic design criteria for Japanese Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, Y.J.; Hofmayer, C.H. [Brookhaven National Lab., Upton, NY (United States); Costello, J.F. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-04-01

    This paper summarizes the results of recent survey studies on the seismic design practice for nuclear power plants in Japan. The seismic design codes and standards for both nuclear as well as non-nuclear structures have been reviewed and summarized. Some key documents for understanding Japanese seismic design criteria are also listed with brief descriptions. The paper highlights the design criteria to determine the seismic demand and component capacity in comparison with U.S. criteria, the background studies which have led to the current Japanese design criteria, and a survey of current research activities. More detailed technical descriptions are presented on the development of Japanese shear wall equations, design requirements for containment structures, and ductility requirements.

  20. Core safety of Indian nuclear power plants (NPPs) under extreme conditions

    Indian Academy of Sciences (India)

    J B Joshi; A K Nayak; M Singhal; D Mukhopadhaya

    2013-10-01

    Nuclear power is currently the fourth largest source of electricity production in India after thermal, hydro and renewable sources of electricity. Currently, India has 20 nuclear reactors in operation and seven other reactors are under construction. Most of these reactors are indigenously designed and built Heavy Water Reactors. In addition, a 300 MWe Advanced Heavy Water Reactor has already been designed and in the process of deployment in near future for demonstration of power production from Thorium apart from enhanced safety features by passive means. India has ambitious plans to enhance the share of electricity production from nuclear. The recent Fukushima accident has raised concerns of safety of Nuclear Power Plants worldwide. The Fukushima accident was caused by extreme events, i.e., large earthquake followed by gigantic Tsunami which are not expected to hit India’s coast considering the geography of India and historical records. Nevertheless, systematic investigations have been conducted by nuclear scientists in India to evaluate the safety of the current Nuclear Power Plants in case of occurrence of such extreme events in any nuclear site. This paper gives a brief outline of the safety features of Indian Heavy Water Reactors for prevention and mitigation of such extreme events. The probabilistic safety analysis revealed that the risk from Indian Heavy Water Reactors are negligibly small.

  1. Analysis of three-phase power transformer laminated magnetic core designs

    Directory of Open Access Journals (Sweden)

    M.I. Levin

    2014-03-01

    Full Text Available Analysis and research into properties and parameters of different-type laminated magnetic cores of three-phase power transformers are conducted. Most of new laminated magnetic core designs are found to have significant shortcomings resulted from design and technological features of their manufacturing. These shortcomings cause increase in ohmic loss in the magnetic core, which eliminates advantages of the new core configurations and makes them uncompetitive as compared with the classical laminated magnetic core design.

  2. Safeguards-by-Design: Early Integration of Physical Protection and Safeguardability into Design of Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    T. Bjornard; R. Bean; S. DeMuth; P. Durst; M. Ehinger; M. Golay; D. Hebditch; J. Hockert; J. Morgan

    2009-09-01

    The application of a Safeguards-by-Design (SBD) process for new nuclear facilities has the potential to minimize proliferation and security risks as the use of nuclear energy expands worldwide. This paper defines a generic SBD process and its incorporation from early design phases into existing design / construction processes and develops a framework that can guide its institutionalization. SBD could be a basis for a new international norm and standard process for nuclear facility design. This work is part of the U.S. DOE’s Next Generation Safeguards Initiative (NGSI), and is jointly sponsored by the Offices of Non-proliferation and Nuclear Energy.

  3. 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...

  4. 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

  5. Design and operation of the core topography data acquisition system for TMI-2

    Energy Technology Data Exchange (ETDEWEB)

    Beller, L S; Brown, H L

    1984-05-01

    Development of effective procedures for recovery from the 1979 accident at the Three Mile Island 2 nuclear station requires a detailed and quantitative description of the postaccident configuration of the core. This report describes the techniques, equipment, and procedures used for making precise ultrasonic, sonar-like measurements of the cavity left in the upper core region as a result of the accident and details the primary results of the measurements. The system developed for the measurements uses computer techniques for the command and control of remote mechanical and electronic equipment, and for data acquisition and reduction. The system was designed, fabricated, and tested; procedures developed; and personnel trained in 4-1/2 months. The primary results are detailed topographic maps of the cavity. A variety of visual aids was developed to supplement the maps and aid in interpreting companion videotape surveys. The measurements reveal a cavity of 9.3 m/sup 3/, approximately 26% of the total core volume. The cavity occupies most of the full diameter of the core to an average depth of about 1.5 m and approaches 2 m in places.

  6. The Next Generation Nuclear Plant - Insights Gained from the INEEL Point Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald; A. M. Baxter; P. D. Bayless; J. M. Bolin; H. D. Gougar; R. L. Moore; A. M. Ougouag; M. B. Richards; R. L. Sant; J. W. Sterbentz; W. K. Terry

    2004-08-01

    This paper provides the results of an assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebble-bed fuel helium gas reactor. Insights gained regarding the strengths and weaknesses of the two designs are also discussed. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. Two major modifications of the current Gas Turbine- Modular Helium Reactor (GT-MHR) design were needed to obtain a prismatic block design with a 1000 C outlet temperature: reducing the bypass flow and better controlling the inlet coolant flow distribution to the core. The total power that could be obtained for different core heights without exceeding a peak transient fuel temperature of 1600 °C during a high or low-pressure conduction cooldown event was calculated. With a coolant inlet temperature of 490 °C and 10% nominal core bypass flow, it is estimated that the peak power for a 10-block high core is 686 MWt, for a 12-block high core is 786 MWt, and for a 14-block core is about 889 MWt. The core neutronics calculations showed that the NGNP will exhibit strongly negative Doppler and isothermal temperature coefficients of reactivity over the burnup cycle. In the event of rapid loss of the helium gas, there is negligible core reactivity change. However, water or steam ingress into the core coolant channels can produce a relatively large reactivity effect. Two versions of an annular pebble-bed NGNP have also been developed, a 300 and a 600 MWt module. From this work we learned how to design passively safe pebble bed reactors that produce more than 600 MWt. We also found a way to improve both the fuel utilization and safety by modifying the pebble design (by adjusting the fuel zone radius in the pebble to optimize the fuel

  7. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  8. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  9. Supporting Common Core Instruction With Literacy Design Collaborative

    Directory of Open Access Journals (Sweden)

    Joan Herman

    2016-06-01

    Full Text Available The article examines the results of two quasi-experimental studies of the implementation and impact of the Literacy Design Collaborative (LDC, an intervention designed to support secondary teachers’ transition to Common Core State Standards in English language arts. The first study examines LDC implementation by eighth-grade social studies and science teachers in districts across Kentucky; the second study is set in sixth-grade advanced reading classes in a large urban district in Florida. Based on teacher surveys, logs, and analysis of classroom artifacts, the LDC was implemented with reasonable fidelity across both studies. Based on available assessment scores, results show statistically significant positive effects in Kentucky for reading but no corollary effect in Florida. There were no significant differences in writing scores in either site. The conclusion shares hypotheses that may explain the differences in results and reflects on implications for evidence-based practice.

  10. The synthesis, design and applications of lanthanide cored complexes

    Science.gov (United States)

    Phelan, Gregory David

    Novel luminescent materials based on lanthanide cored complexes have been designed and synthesized. The complexes consist of a beta-diketone ligand chelated to a lanthanide metal such as europium or gadolinium. A series of beta-diketone ligands were designed and synthesized. The ligands consist of a polycyclic aromatic sensitizer, phenanthrene, and a second functional group. The second groups consisted of another unit of phenanthrene, a dendritic structure, or a fluorinated alkyl chain. The europium complexes have been incorporated into organic light emitting devices that have a major emission at 615 nm and a maximum brightness of 300 cd/m2. The gadolinium complexes were used to dope into the resulting organic light emitting devices to help improve the efficiency of the device. The use of the gadolinium complexes results in a 25 fold increase in efficiency.

  11. Optimizing the Design of Small Fast Spectrum Battery-Type Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Staffan Qvist

    2014-07-01

    Full Text Available This study is focused on defining and optimizing the design parameters of inherently safe “battery” type sodium-cooled metallic-fueled nuclear reactor cores that operate on a single stationary fuel loading at full power for 30 years. A total of 29 core designs were developed with varying power and flow conditions, including detailed thermal-hydraulic, structural-mechanical and neutronic analysis. Given set constraints for irradiation damage, primary cycle pressure drop and inherent safety considerations, the attainable power range and performance characteristics of the systems are defined. The optimum power level for a core with a coolant pressure drop limit of 100 kPa and an irradiation damage limit of 200 DPA (displacements per atom is found to be 100 MWt/40 MWe. Raising the power level of an optimized core gives significantly higher attainable power densities and burnup, but severely decreases safety margins and increases the irradiation damage. A fully optimized inherently safe battery-type fast reactor core with an active height and diameter of 150 cm (2.6 m3, a pressure drop limit of 100 kPa and an irradiation damage limit of 300 DPA can be designed to operate at 150 MWt/60 MWe for 30 years, reaching an average discharge burnup of 100 MWd/kg-actinide.

  12. Design Basis of Core Components and their Realization in the frame of the EPR's{sup TM} Core Component Development

    Energy Technology Data Exchange (ETDEWEB)

    Schebitz, Florian [AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Mekmouche, Abdelhalim [AREVA NP SAS, 10 rue Juliette Recamier, 69456 Lyon Cedex 06 (France)

    2008-07-01

    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{sup 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{sup TM}, the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR{sup TM}. (authors)

  13. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  14. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  15. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    Energy Technology Data Exchange (ETDEWEB)

    Meneses, Anderson Alvarenga de Moura, E-mail: ameneses@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program; Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Gambardella, Luca Maria, E-mail: luca@idsia.c [Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Schirru, Roberto, E-mail: schirru@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

    2009-07-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  16. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

  17. Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

    Science.gov (United States)

    Hsieh, T.-M.; Koenig, D. R.

    1977-01-01

    Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

  18. Nuclear structure and the fate of core collapse (Type II) supernova

    Energy Technology Data Exchange (ETDEWEB)

    Gai, Moshe [LNS at Avery Point, University of Connecticut, Groton, CT 06340-6097 (United States); Wright Lab, Dept. of Physics, Yale University, New Haven, CT 06520-8124 (United States)

    2014-08-15

    For a long time Gerry Brown and his collaborator Hans Bethe considered the question of the final fate of a core collapse (Type II) supernova. Recalling ideas from nuclear structure on Kaon condensate and a soft equation of state of the dense nuclear matter they concluded that progenitor stars with mass as low as 17–18M{sub ⊙} (including supernova 1987A) could collapse to a small mass black hole with a mass just beyond 1.5M{sub ⊙}, the upper bound they derive for a neutron star. We discuss another nuclear structure effect that determines the carbon to oxygen ratio (C/O) at the end of helium burning. This ratio also determines the fate of a Type II supernova with a carbon rich progenitor star producing a neutron star and oxygen rich collapsing to a black hole. While the C/O ratio is one of the most important nuclear inputs to stellar evolution it is still not known with sufficient accuracy. We discuss future efforts to measure with gamma-beam and TPC detector of the {sup 12}C(α,γ){sup 16}O reaction that determines the C/O ratio in stellar helium burning.

  19. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    Science.gov (United States)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  20. Nuclear Structure and the Fate of Core Collapse (Type II) Supernova

    CERN Document Server

    Gai, Moshe

    2014-01-01

    For a long time Gerry Brown and his collaborator Hans Bethe considered the question of the final fate of a core collapse (Type II) supernova. Recalling ideas from nuclear structure on Kaon condensate and a soft equation of state of the dense nuclear matter they concluded that progenitor stars with mass as low a 17-18M$_\\odot$ (including supernova 1987A) could collapse to a small mass black hole with a mass just beyond 1.5M$_\\odot$, the upper bound they derive for a neutron star. We discuss another nuclear structure effect that determines the carbon to oxygen ratio (C/O) at the end of helium burning. This ratio also determines the fate of a Type II supernova with a carbon rich progenitor star producing a neutron star and oxygen rich collapsing to a black hole. While the C/O ratio is one of the most important nuclear input to stellar evolution it is still not known with sufficient accuracy. We discuss future efforts to measure with gamma-beam and TPC detector the 12C(a,g)16O reaction that determines the C/O rat...

  1. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  2. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems

    Energy Technology Data Exchange (ETDEWEB)

    Gallup, D.R.

    1990-03-01

    In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

  3. An out-of-core thermionic-converter system for nuclear space power.

    Science.gov (United States)

    Breitwieser, R.

    1972-01-01

    Reexamination of designs of nuclear thermionic space power systems with the converter outside the reactor in the perspective of recent advances in heat-transfer methods, materials, converter performance, and radiation design. The 40- to 70-kW(e) power range is treated. The configuration is found to meet the constraints of readily available launch vehicles. It allows for off-design operation including startup, shutdown, and possible emergency conditions; provides tolerance of failure by extensive use of modular, redundant elements; incorporates and uses heat pipes in a fashion that reduces the need for extensive in-pile testing of system components; and uses thermionic converters, nuclear fuel elements, and heat-transfer devices in a geometrical form adapted from existing incore thermionic system designs.

  4. Gas core reactor power plants designed for low proliferation potential

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    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/sub 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 /sup 233/U born from thorium. Fission product removal was continuous. Newly born /sup 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 /sup 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.

  5. Dynamico, an Icosahedral Dynamical Core Designed for Consistency and Versatility

    Science.gov (United States)

    Dubos, T.

    2014-12-01

    The design of the icosahedral-hexagonal dynamical core DYNAMICO is presented. DYNAMICO solves the multi-layer rotating shallow-water equations, a compressible variant of the same equivalent to a discretization of the hydrostatic primitive equations (HPE) in a Lagrangian vertical coordinate, and the HPE in a hybrid mass-based vertical coordinate. In line with more general lines of thought known as physics-preserving discretizations and discrete differential geometry, kinematics and dynamics are separated as strictly as possible. This separation means that the transport of mass, scalars and potential temperature uses no information regarding the specific momentum equation being solved. This disregarded information includes the equation of state as well as any metric information, and is used only for certain terms of the momentum budget, written in Hamiltonian, vector-invariant form. The common Hamiltonian structure of the various equations of motion (Tort and Dubos, 2014 ; Dubos and Tort, 2014) is exploited to formulate energy-conserving spatial discretizations in a unified way. Furthermore most of the model code is common to the three sets of equations solved, making it easier to develop and validate each piece of the model separately. This design permits to consider several extensions in the near future, especially to deep-atmosphere, moist and non-hydrostatic equations. Representative academic three-dimensional benchmarks are run and analyzed, showing correctness of the model (Figure : time-zonal statistics from Held and Suarez (1994) simulations). Hopefully preliminary full-physics results will be presented as well. References : T. Dubos and M. Tort, "Equations of atmospheric motion in non-Eulerian vertical coordinates : vector-invariant form and Hamiltonian formulation", accepted by Mon. Wea. Rev. M. Tort and T. Dubos, "Usual approximations to the equations of atmospheric motion : a variational perspective" accepted by J. Atmos. Sci T. Dubos et al., "DYNAMICO

  6. 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.

  7. Fuel management strategy for the new equilibrium silicide core design of RSG GAS (MPR-30)

    Energy Technology Data Exchange (ETDEWEB)

    Hong Liem Peng; Arbie, Bakri; Sembiring, T.M. [National Atomic Energy Agency (Batan), Center for Multipurpose Reactor, Tangerang (Indonesia)

    1997-07-01

    The design procedure and fuel management strategy were proposed for converting the oxide core of RSG GAS (MPR-30) to the new equilibrium silicide core using higher uranium loading. The obtained silicide core gave significant extension of the core cycle length and thus increasing the reactor availability and utilisation. (author)

  8. Fuel management strategy for the new equilibrium silicide core design of RSG GAS (MPR-30)

    Energy Technology Data Exchange (ETDEWEB)

    Hong Liem Peng; Arbie, Bakri; Sembiring, T.M. [National Atomic Energy Agency (Batan), Center for Multipurpose Reactor, Tangerang (Indonesia)

    1997-07-01

    The design procedure and fuel management strategy were proposed for converting the oxide core of RSG GAS (MPR-30) to the new equilibrium silicide core using higher uranium loading. The obtained silicide core gave significant extension of the core cycle length and thus increasing the reactor availability and utilisation. (author) 4 figs., 1 tab., refs.

  9. Design of homogeneous trench-assisted multi-core fibers based on analytical model

    DEFF Research Database (Denmark)

    Ye, Feihong; Tu, Jiajing; Saitoh, Kunimasa

    2016-01-01

    is the quasi-optimum core layout starting from an one-ring structured 12-core fiber. Based on the analytical model, a square-lattice structured 24-core fiber and a 32-core fiber are designed both for propagation-direction interleaving (PDI) and non-PDI transmission schemes. The proposed model provides...

  10. SPOUTED BED DESIGN CONSIDERATIONS FOR COATED NUCLEAR FUEL PARTICLES

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas W.

    2017-07-01

    High Temperature Gas Cooled Reactors (HTGRs) are fueled with tristructural isotropic (TRISO) coated nuclear fuel particles embedded in a carbon-graphite fuel body. TRISO coatings consist of four layers of pyrolytic carbon and silicon carbide that are deposited on uranium ceramic fuel kernels (350µm – 500µm diameters) in a concatenated series of batch depositions. Each layer has dedicated functions such that the finished fuel particle has its own integral containment to minimize and control the release of fission products into the fuel body and reactor core. The TRISO coatings are the primary containment structure in the HTGR reactor and must have very high uniformity and integrity. To ensure high quality TRISO coatings, the four layers are deposited by chemical vapor deposition (CVD) using high purity precursors and are applied in a concatenated succession of batch operations before the finished product is unloaded from the coating furnace. These depositions take place at temperatures ranging from 1230°C to 1550°C and use three different gas compositions, while the fuel particle diameters double, their density drops from 11.1 g/cm3 to 3.0 g/cm3, and the bed volume increases more than 8-fold. All this is accomplished without the aid of sight ports or internal instrumentation that could cause chemical contamination within the layers or mechanical damage to thin layers in the early stages of each layer deposition. The converging section of the furnace retort was specifically designed to prevent bed stagnation that would lead to unacceptably high defect fractions and facilitate bed circulation to avoid large variability in coating layer dimensions and properties. The gas injection nozzle was designed to protect precursor gases from becoming overheated prior to injection, to induce bed spouting and preclude bed stagnation in the bottom of the retort. Furthermore, the retort and injection nozzle designs minimize buildup of pyrocarbon and silicon carbide on the

  11. Fully encapsulated directional self-powered gamma ray detector for use in in-core nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    LeVert, F.E.; Cox, S.A.

    1979-01-01

    A study of a fully encapsulated directional self-powered gamma ray detector designed for localized in core measurements in a nuclear reactor was conducted. The detector consisted of a multilayer arrangement of a metal-dielectric-metal-dielectric-metal structure. The dielectric material was made of two plates of unequal thicknesses which were placed on opposite sides of the central metal plate. The direction discrimination exhibited by the detector was attributed to the combined effect of electron ranges, Photo-Compton electron generation rates, and the presence of E-fields in the unequal thicknesses of dielectric material. Results showing the response of the detector when it was placed in a gamma ray field with a known anisotropic component are presented.

  12. Perspectives on Validation and Uncertainty Evaluation of SFR Nuclear Design Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moohoon; Choi, Yong Won; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    Fast reactors such as PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) developed by KAERI have fundamental differences in terms of core characteristics and associated fuel cycle compared to thermal reactors, which need specific new effort for code validation. In current PWRs, nuclear design code systems have been validated using numerous data accumulated by wide operating experience, and its uncertainty can be assessed by statistical methods. However, in order to validate code systems for SFRs with little operating experience, and particularly prototype reactor, new approaches are required. In this study, a current procedure for validation and uncertainty evaluation is reviewed in nuclear design code systems for PWRs, and global approaches for validation of SFR code systems are surveyed. Through these reviews, perspectives on nuclear design code validation for SFRs are identified. In case of neutronics code V and V, current procedure for PWRs and global approaches for SFRs were reviewed and surveyed. Though this review, perspectives on nuclear design code V and V and uncertainty evaluation for SFRs were identified. Further study will be implemented to obtain more insight on code validation.

  13. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  14. Conceptual design study on an upgraded future Monju core (2). Core concept with extended refueling interval and increased fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kinjo, Hidehito; Ishibashi, Jun-ichi; Nishi, Hiroshi [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Kageyama, Takeshi [Nuclear Energy System Inc., Tokyo (Japan)

    2003-03-01

    A conceptual design study has been performed at the International Cooperation and Technology Development Center to investigate the feasibility of upgraded future Monju cores with extended refueling intervals of 365efpd/cycle and increased fuel burnup of 150 GWd/t. The goal of this study is to demonstrate the possible contribution of Monju to the improved economy and to efficient utilization, as one of the major facilities for fast neutron irradiation. Two design measures have been mainly taken to improve the core fuel burnup and reactivity control characteristics for the extended operating cycle length of 1 year: (1) The driver fuel pin specification with both increased pin diameter of 7.7mm and increased active core height of about 100cm has been chosen to reduce the burnup reactivity swing, (2) The absorber control rod specification has also been changed to enhance the control rod reactivity worth by increasing {sup 10}B-enrichment and absorber length, and to adequately secure the shutdown reactivity margin. The major core characteristics have been evaluated on the core power distribution, safety parameters such as sodium void reactivity and Doppler effect, thermal hydraulics and reactivity control characteristics. The results show that this core could achieve the targeted core performances of 1-year operating cycle as well as 150GWd/t discharged burnup, without causing any significant drawback on the core characteristics and safety aspects. The upgraded core concepts have, therefore, been confirmed as feasible. (author)

  15. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  16. Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

    2017-09-01

    The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

  17. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Energy Technology Data Exchange (ETDEWEB)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki [Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan) and Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

    2012-11-12

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  18. Possible generation of heat from nuclear fusion in Earth’s inner core

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-01

    The cause and source of the heat released from Earth’s interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2D + 2D + 2D → 21H + 4He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 1012 J/m3, based on the assumption that Earth’s primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth’s interior to the universe, and pass through Earth, respectively.

  19. Possible generation of heat from nuclear fusion in Earth's inner core.

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-23

    The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: (2)D + (2)D + (2)D → 2(1)H + (4)He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10(12) J/m(3), based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

  20. Design rules for core/shell nanowire resonant emitters

    Science.gov (United States)

    Kim, Da-Som; Kim, Sun-Kyung

    2017-01-01

    We study design principles to boost the extraction of light from core/shell GaN nanowire optical emitters. A full-vectorial electromagnetic simulation reveals that the extraction efficiency of an emitter within a nanowire cavity depends strongly on its position; the efficiency becomes maximized as the emitter's location approaches the center of the structure. The total extraction of light is sinusoidally modulated by the nanowire diameter, which is directly correlated with optical resonances. The introduction of a conformal dielectric coating on a nanowire leads to a dramatic enhancement in the extraction efficiency, which results from an increase in side emission owing to an optical antenna effect. A simple high-refractive-index dielectric coating approximately doubles the total extraction efficiency of a nanowire LED. These numerical findings will be valuable in providing strategies for high-efficiency nanowire-based optical emitters.

  1. Upgrading of seismic design of nuclear power plant building

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi [Tokyo Univ. (Japan). Faculty of Engineering; Kitada, Yoshio

    1997-03-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan`s effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  2. Safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2002-04-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term nuclear R and D Program. In this report, key safety design features are described and safety analyses results for typical ATWS accidents in the KALIMER design with breakeven core are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the event categorization and acceptance criteria for the KALIMER safety analysis are described in chapter 2. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER system response to the events. They are categorized as Bounding Events (BEs) because of their low probability of occurrence. In chapter 4, the performance analysis results of the KALIMER containment dome are described along with the HCDA accident scenario and source terms. The major containment parameters of peak pressure and peak temperature have been calculated using the CONTAIN-LMR code. Radiological consequence has been evaluated by the MACCS code. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using SCHAMBETA code developed in the framework of the modified bethe-tait method. Work energy potentials based arising from the sodium expansion as well as the isentropic fuel expansion are then calculated to evaluate the structural integrity of the reactor vessel, reactor internals and primary coolant system of KALIMER.

  3. The design of reload cores using optimal control theory

    Energy Technology Data Exchange (ETDEWEB)

    Terney, W.B.; Williamson, E.A.

    1982-11-01

    A formal approach for the optimization of the final design of reload cores has been devised and verified. The method is based on applying the calculus of variations (Pontryagin's principle) to the normal flux and depletion system equations. The resulting set of coupled system, Euler-Lagrange (E-L), and optimality equations are solved iteratively. This is done by assuming a loading pattern for the old fuel, first solving the system equations, and then the E-L equations. The pattern is then modified by using the optimality (or Pontryagin) condition, and the process is repeated until no further improvements can be made. A computer program, OPMUV, implementing these procedures has been written and verified. The code can handle two-dimensional, quarter-core symmetric configurations with up to 241 assemblies and 4 nodes per assembly with modified one-group theory. It also has the capability of optimizing over the entire depletion cycle as well as just at the beginning of cycle (BOC). The results show that the procedure does work. In all cases tried, the method led to a reduction in nodal peaks of 1 to 3% over the final designer-obtained loading pattern within a couple of iterations. These savings carry over to comparable reductions in pin peaks when the optimized patterns are used in four-group, fine-mesh calculations. Since the changes on each iteration are limited to ensure convergence, the method is thus well suited for the final fine tuning of the normally obtained patterns to gain an extra few percent in power flattening.

  4. Parallel VLSI design for the fast -D DWT core algorithm

    Institute of Scientific and Technical Information of China (English)

    WEI Benjie; LIU Mingye; ZHOU Yihua; CHENG Baodong

    2007-01-01

    By studying the core algorithm of a three-dimensional discrete wavelet transform (3-D DWT) in depth,this Paper divides it into three one-dimensional discrete wavelet transforms (1-D DWTs).Based on the implementation of a 3-D DWT software,a parallel architecture design of a very large-scale integration(VLSI)is produced.It needs three dual-port random-access memory(RAM)to store the temporary results and transpose the matrix,then builds up a pipeline model composed of the three 1-D DWTs.In the design.the finite state machine(FSM)is used well to control the flow.Compared with the serial mode.the experimental results of the post synthesized simulation show that the design method is correct and effective.It can increase the processing speed by about 66%.work at 59 MHz,and meet the real-time needs of the video encoder.

  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. AP1000, a nuclear central of advanced design; AP1000, una central nuclear de diseno avanzado

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N.; Viais J, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: nhm@nuclear.inin.mx

    2005-07-01

    The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

  7. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  8. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Jonathan, E-mail: JMitchell16@slb.com; Fordham, Edmund J. [Schlumberger Gould Research, High Cross, Madingley Road, Cambridge CB3 0EL (United Kingdom)

    2014-11-15

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general.

  9. Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

    Science.gov (United States)

    Viellieber, Mathias; Class, Andreas G.

    2013-11-01

    Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

  10. Contributed review: nuclear magnetic resonance core analysis at 0.3 T.

    Science.gov (United States)

    Mitchell, Jonathan; Fordham, Edmund J

    2014-11-01

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general.

  11. Multimegawatt nuclear electric propulsion with gaseous and vapor core reactors with MHD

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim; Smith, Blair; Houts, Michael

    2001-02-01

    This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a fissioning plasma core reactor (FPCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasmadynamic (MPD) thruster. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Candidate working fluids include K, Li, Na, KF, LiF, NaF, etc. The system features core outlet temperatures of 3000 to 4000 K at pressures of about 1 to 10 MPa, MHD temperatures of 2000 to 3000 K, and radiator temperatures of 1200 to 2000 K. This combination of parameters offers the potential for low total system specific mass in the range of 0.4 to 0.6 kg/kWe. The MHD output could be coupled with minimal power conditioning to the variable specific impulse magnetoplasma rocket (VASIMR), MPD thrusters or other types of thruster for producing thrust at very high specific impulse (Isp=1500 to 10,000 s). .

  12. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  13. Nuclear inputs of key iron isotopes for core-collapse modeling and simulation

    CERN Document Server

    Nabi, Jameel-Un

    2014-01-01

    From the modeling and simulation results of presupernova evolution of massive stars, it was found that isotopes of iron, $^{54,55,56}$Fe, play a significant role inside the stellar cores, primarily decreasing the electron-to-baryon ratio ($Y_{e}$) mainly via electron capture processes thereby reducing the pressure support. The neutrinos produced, as a result of these capture processes, are transparent to the stellar matter and assist in cooling the core thereby reducing the entropy. The structure of the presupernova star is altered both by the changes in $Y_{e}$ and the entropy of the core material. Here we present the microscopic calculation of Gamow-Teller strength distributions for isotopes of iron. The calculation is also compared with other theoretical models and experimental data. Presented also are stellar electron capture rates and associated neutrino cooling rates, due to isotopes of iron, in a form suitable for simulation and modeling codes. It is hoped that the nuclear inputs presented here should ...

  14. Nuclear design report for system-integrated modular advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Lee, Chung Chan; Zee, Sung Quun; Chang, Moon Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-02-01

    This report presents nuclear characteristics analysis results for SMART. Information is given on fuel loading, power density distributions, reactivity coefficients and control rod worths. The core consists of 57 modified Korean Standard Fuel Assemblies (m-KOFAs). and all fuel assemblies contain burnable absorbers to control the power distribution and the excess reactivity that is required for soluble boron-free and ultra longer cycle operation. The cycle length of SMART amounts to 990 EFPD corresponding to a cycle burnup of 26,160 MWD/MTU. 4 refs., 92 figs., 5 tabs. (Author)

  15. Tokamak Fusion Core Experiment: design studies based on superconducting and hybrid toroidal field coils. Design overview

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, C.A. (ed.)

    1984-10-01

    This document is a design overview that describes the scoping studies and preconceptual design effort performed in FY 1983 on the Tokamak Fusion Core Experiment (TFCX) class of device. These studies focussed on devices with all-superconducting toroidal field (TF) coils and on devices with superconducting TF coils supplemented with copper TF coil inserts located in the bore of the TF coils in the shield region. Each class of device is designed to satisfy the mission of ignition and long pulse equilibrium burn. Typical design parameters are: major radius = 3.75 m, minor radius = 1.0 m, field on axis = 4.5 T, plasma current = 7.0 MA. These designs relay on lower hybrid (LHRH) current rampup and heating to ignition using ion cyclotron range of frequency (ICRF). A pumped limiter has been assumed for impurity control. The present document is a design overview; a more detailed design description is contained in a companion document.

  16. Spent nuclear fuel canister storage building conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  17. SCW Pressure-Channel Nuclear Reactor Some Design Features

    Science.gov (United States)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  18. Identifying and Using ‘Core Competencies’ to Help Design and Assess Undergraduate Neuroscience Curricula

    OpenAIRE

    Kerchner, Michael; Hardwick, Jean C.; Thornton, Janice E.

    2012-01-01

    There has been a growing emphasis on the use of core competencies to design and inform curricula. Based on our Faculty for Undergraduate Neuroscience workshop at Pomona we developed a set of neuroscience core competencies. Following the workshop, faculty members were asked to complete an online survey to determine which core competencies are considered most essential and the results are presented. Backward Design principles are then described and we discuss how core competencies, through a ba...

  19. 76 FR 63541 - Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants

    Science.gov (United States)

    2011-10-13

    ...-2010-0288] Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants AGENCY: Nuclear... Hurricane Missiles for Nuclear Power Plants.'' This regulatory guide provides licensees and applicants with... hurricane and design-basis hurricane-generated missiles that a nuclear power plant should be designed...

  20. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  1. Development of Reactor Core for Nuclear Thermal Propulsion%核热推进堆芯方案的发展

    Institute of Scientific and Technical Information of China (English)

    解家春; 赵守智

    2012-01-01

    Nuclear thermal propulsion heats propellant with fission energy. It's specific impulse is double of chemical rockets. It could play an important role in space mission. During the research process about nuclear thermal propulsion in USA and Russia, many reactors were well developed. The details of the reactors core were described, the characteristics of design were indicated, and the trend of development was summarized.%核热推进利用核裂变能加热工质,比冲可达化学火箭的2倍多,在空间活动中有广阔的应用前景.在美国和俄罗斯的研究过程中,对多个核热推进堆芯方案进行了较深入的研究.本工作介绍了这些堆芯方案的情况,详细说明了其设计特点,并总结了堆芯方案的发展趋势.

  2. A design study of sodium cooled metal fuel core for high outlet-temperature

    Energy Technology Data Exchange (ETDEWEB)

    Yamadate, Megumi; Mizuno, Tomoyasu; Sugino, Kazuteru [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2003-03-01

    A design study of sodium cooled metal fuel core was performed. The new core concept studied here has low radial power peaking by applying three regions core configuration with different diameters of fuel pins and the same enrichment of plutonium. The core reveals constant radial power distribution during nominal power operation, which gives the advantage of low cladding maximum temperature or high core outlet temperature with a cladding limit temperature. The core thermal hydraulic design shows that a core outlet temperature as high as that of the oxide fuel core is feasible even in the application of metal fuel pins, which have a lower cladding limit temperature than oxide fuel pins. The core concept is possible to have additional attractiveness such as high breeding ratio, high burnup, and long-term operation cycle due to its high internal conversion ratio. (author)

  3. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR) standard...

  4. Nuclear-powered Hysat spacecraft: comparative design study

    Energy Technology Data Exchange (ETDEWEB)

    Raab, B.

    1975-08-01

    The study shows that the all-nuclear spacecraft can have a substantial weight advantage over a hybrid (nuclear/solar) or all-solar spacecraft, owing to a further reduction in power requirement, and to the elimination of such equipment as the sensor gimbal and rotating joint assemblies. Because the need for a sun-oriented section is eliminated, the all-nuclear spacecraft can be designed as a monolithic structure, with the sensor and other payload firmly secured in a fixed position on the structure. This enhances attitude stability while minimizing structural weight and eliminating the need for flexible fluid lines. Sensor motion can be produced, varied, and controlled within the limits specified by the study contractors by moving the entire spacecraft in the prescribed pattern. A simple attitude control system using available hardware suffices to meet all requirements.

  5. Basic Research and Development Effort to Design a Micro Nuclear Power Plant for Brazilian Space Applications

    Science.gov (United States)

    Guimares, L. N. F.; Camillo, G. P.; Placco, G. M.; Barrios, G., A., Jr.; Do Nascimento, J. A.; Borges, E. M.; De Castro Lobo, P. D.

    For some years the Nuclear Energy Division of the Institute for Advanced Studies is conducting the TERRA (Portuguese abbreviation for advanced fast reactor technology) project. This project aims at research and development of the key issues related with nuclear energy applied to space technology. The purpose of this development is to allow future Brazilian space explorers the access of a good and reliable heat, power and/or propulsion system based on nuclear energy. Efforts are being made in fuel and nuclear core design, designing and building a closed Brayton cycle loop for energy conversion, heat pipe systems research for passive space heat rejection, developing computational programs for thermal loop safety analysis and other technology that may be used to improve efficiency and operation. Currently there is no specific mission that requires these technology development efforts; therefore, there is a certain degree of freedom in the organization and development efforts. This paper will present what has been achieved so far, what is the current development status, where efforts are heading and a proposed time table to meet development objectives.

  6. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  7. Assessment of water hammer effects on boiling water nuclear reactor core dynamics

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2007-01-01

    Full Text Available Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

  8. Assessing the feasibility and consequences of nuclear georeactors in the Earths core mantle boundary

    CERN Document Server

    De Meijer, R J

    2008-01-01

    We assess the likelihood and geochemical consequences of the presence of nuclear georeactors in the core mantle boundary region (CMB) between Earths silicate mantle and metallic core. Current geochemical models for the Earths interior predict that U and Th in the CMB are concentrated exclusively in the mineral calcium silicate perovskite (CaPv), leading to predicted concentration levels of approximately 12 ppm combined U and Th, 4.5 Ga ago if CaPv is distributed evenly throughout the CMB. Assuming a similar behaviour for primordial 244Pu provides a considerable flux of neutrons from spontaneous fission. We show that an additional concentration factor of only an order of magnitude is required to both ignite and maintain self sustaining georeactors based on fast fission. Continuously operating georeactors with a power of 5 TW can explain the observed isotopic compositions of helium and xenon in the Earths mantle. Our hypothesis requires the presence of elevated concentrations of U and Th in the CMB, and is amen...

  9. Nanostructure Core Fiber With Enhanced Performances: Design, Fabrication and Devices

    DEFF Research Database (Denmark)

    Yu, X.; Yan, Min; Ren, G.B.;

    2009-01-01

    We report a new type of silica-based all-solid fiber with a 2-D nanostructure core. The nanostructure core fiber (NCF) is formed by a 2-D array of high-index rods of sub-wavelength dimensions. We theoretically study the birefringence property of such fibers over a large wavelength range. Large...

  10. 78 FR 32988 - Core Principles and Other Requirements for Designated Contract Markets; Correction

    Science.gov (United States)

    2013-06-03

    ... COMMISSION 17 CFR Part 38 RIN 3038-AD09 Core Principles and Other Requirements for Designated Contract...: This document corrects the Federal Register release of the final rule regarding Core Principles and... language for the previously published Federal Register release of the final rule regarding Core...

  11. 76 FR 14825 - Core Principles and Other Requirements for Designated Contact Markets

    Science.gov (United States)

    2011-03-18

    ... COMMISSION 17 CFR Parts 1, 16, and 38 RIN 3038-AD09 Core Principles and Other Requirements for Designated... Commission in the Federal Register release for the notice of proposed rulemaking for ``Core Principles and... comment period for the proposed rulemaking closed on February 22, 2011. \\2\\ See Core Principles and...

  12. 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)

  13. Novel design of hollow-core multi clad fiber for long haul optical communication system

    Science.gov (United States)

    Palodiya, Vikram; Raghuwanshi, Sanjeev K.

    2016-09-01

    We have described a dispersion characteristics of hollow-core multi-clad index profiles, which include a hollow core. The designs satisfy the most important requirements for applications in long haul communication. This design fiber shows zero dispersion at 1550 nm can be obtained for the fundamental air core mode over a wide wavelength range by introducing the partial reflector layer around the core, optimizing expanded core size and silica cladding thickness. Also analyze dispersion compensating properties of these fibers. This unique structure of the fundamental air core mode is presented by the introduction of partial reflector cladding around the core. The potential applications of hollow-core multi clad fibers in long-haul optical communication system.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  15. A new equation of state for core-collapse supernovae based on realistic nuclear forces and including a full nuclear ensemble

    Science.gov (United States)

    Furusawa, S.; Togashi, H.; Nagakura, H.; Sumiyoshi, K.; Yamada, S.; Suzuki, H.; Takano, M.

    2017-09-01

    We have constructed a nuclear equation of state (EOS) that includes a full nuclear ensemble for use in core-collapse supernova simulations. It is based on the EOS for uniform nuclear matter that two of the authors derived recently, applying a variational method to realistic two- and three-body nuclear forces. We have extended the liquid drop model of heavy nuclei, utilizing the mass formula that accounts for the dependences of bulk, surface, Coulomb and shell energies on density and/or temperature. As for light nuclei, we employ a quantum-theoretical mass evaluation, which incorporates the Pauli- and self-energy shifts. In addition to realistic nuclear forces, the inclusion of in-medium effects on the full ensemble of nuclei makes the new EOS one of the most realistic EOSs, which covers a wide range of density, temperature and proton fraction that supernova simulations normally encounter. We make comparisons with the FYSS EOS, which is based on the same formulation for the nuclear ensemble but adopts the relativistic mean field theory with the TM1 parameter set for uniform nuclear matter. The new EOS is softer than the FYSS EOS around and above nuclear saturation densities. We find that neutron-rich nuclei with small mass numbers are more abundant in the new EOS than in the FYSS EOS because of the larger saturation densities and smaller symmetry energy of nuclei in the former. We apply the two EOSs to 1D supernova simulations and find that the new EOS gives lower electron fractions and higher temperatures in the collapse phase owing to the smaller symmetry energy. As a result, the inner core has smaller masses for the new EOS. It is more compact, on the other hand, due to the softness of the new EOS and bounces at higher densities. It turns out that the shock wave generated by core bounce is a bit stronger initially in the simulation with the new EOS. The ensuing outward propagations of the shock wave in the outer core are very similar in the two simulations, which

  16. Core Power Control of the fast nuclear reactors with estimation of the delayed neutron precursor density using Sliding Mode method

    Energy Technology Data Exchange (ETDEWEB)

    Ansarifar, G.R., E-mail: ghr.ansarifar@ast.ui.ac.ir; Nasrabadi, M.N.; Hassanvand, R.

    2016-01-15

    Highlights: • We present a S.M.C. system based on the S.M.O for control of a fast reactor power. • A S.M.O has been developed to estimate the density of delayed neutron precursor. • The stability analysis has been given by means Lyapunov approach. • The control system is guaranteed to be stable within a large range. • The comparison between S.M.C. and the conventional PID controller has been done. - Abstract: In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a fast nuclear reactor. The reactor core is simulated based on the point kinetics equations and one delayed neutron group. Considering the limitations of the delayed neutron precursor density measurement, a sliding mode observer is designed to estimate it and finally a sliding mode control based on the sliding mode observer is presented. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability.

  17. Comparison of Methods for Evaluating Nuclear Thermal Propulsion Tie-Tube Designs

    Science.gov (United States)

    Kapernick, Richard J.; Dixon, David D.

    2008-01-01

    One of the fundamental structural components in a nuclear thermal rocket design is the tie tube. Proper cooling and flow modeling is important both for the structural integrity of the reactor core and for proper design of downstream components that operate on the hydrogen exiting the tie tube. Two models have been developed. The first is a spreadsheet-based tool designed for sizing tie-tube components, considering mechanical stress and strain limits, deposited moderator power, thermal expansion along the flow path, and conduction from adjacent fuel blocks. The second is a three-dimensional SINDA/FLUINT model used as a benchmark, containing a complete finite-element fuel block and a 1/6th tie-tube model. This paper discusses the performance of both models, as well as the advantages and limitations of each.

  18. Magnet Design Considerations for Fusion Nuclear Science Facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kessel, C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States) Fusion Technology Institute; Titus, P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    2016-06-01

    The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility that provides a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between the International Thermonuclear Experimental Reactor (ITER) and the demonstration power plant (DEMO). Compared with ITER, the FNSF is smaller in size but generates much higher magnetic field, i.e., 30 times higher neutron fluence with three orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5 T at the plasma center with a plasma major radius of 4.8 m and a minor radius of 1.2 m and a peak field of 15.5 T on the toroidal field (TF) coils for the FNSF. Both low-temperature superconductors (LTS) and high-temperature superconductors (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high-performance ternary restacked-rod process Nb3Sn strands for TF magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high-aspect-ratio rectangular CICC design are evaluated for FNSF magnets, but low-activation-jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. The material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.

  19. Seismic design of equipment and piping systems for nuclear power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

    1997-03-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  20. Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

    Science.gov (United States)

    Chen, Shu-cheng S.; Veres, Joseph P.; Fittje, James E.

    2006-01-01

    A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps and turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance characteristics of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed.

  1. Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

    Science.gov (United States)

    Chen, Shu-Cheng S.; Veres, Joseph P.; Fittje, James E.

    2006-01-01

    A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps & turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at the NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance parameters of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed.

  2. Design criteria and realization of a computerized supervisory system for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Maestri, F.; Mangiarotti, M.; Maciocco, G.

    1987-11-01

    This paper describes the design criteria and the realization modalities of a computerized supervisory system for nuclear applications. The man-machine interface design aspects for the Alto Lazio Nuclear Power Plant control room are discussed.

  3. Design process of the nanofluid injection mechanism in nuclear power plants.

    Science.gov (United States)

    Kang, Myoung-Suk; Jee, Changhyun; Park, Sangjun; Bang, In Choel; Heo, Gyunyoung

    2011-04-27

    Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs). This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs) assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD) in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.

  4. Design process of the nanofluid injection mechanism in nuclear power plants

    Directory of Open Access Journals (Sweden)

    Bang In Choel

    2011-01-01

    Full Text Available Abstract Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs. This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.

  5. Sensitivity analysis for reliable design verification of nuclear turbosets

    Energy Technology Data Exchange (ETDEWEB)

    Zentner, Irmela, E-mail: irmela.zentner@edf.f [Lamsid-Laboratory for Mechanics of Aging Industrial Structures, UMR CNRS/EDF, 1, avenue Du General de Gaulle, 92141 Clamart (France); EDF R and D-Structural Mechanics and Acoustics Department, 1, avenue Du General de Gaulle, 92141 Clamart (France); Tarantola, Stefano [Joint Research Centre of the European Commission-Institute for Protection and Security of the Citizen, T.P. 361, 21027 Ispra (Italy); Rocquigny, E. de [Ecole Centrale Paris-Applied Mathematics and Systems Department (MAS), Grande Voie des Vignes, 92 295 Chatenay-Malabry (France)

    2011-03-15

    In this paper, we present an application of sensitivity analysis for design verification of nuclear turbosets. Before the acquisition of a turbogenerator, energy power operators perform independent design assessment in order to assure safe operating conditions of the new machine in its environment. Variables of interest are related to the vibration behaviour of the machine: its eigenfrequencies and dynamic sensitivity to unbalance. In the framework of design verification, epistemic uncertainties are preponderant. This lack of knowledge is due to inexistent or imprecise information about the design as well as to interaction of the rotating machinery with supporting and sub-structures. Sensitivity analysis enables the analyst to rank sources of uncertainty with respect to their importance and, possibly, to screen out insignificant sources of uncertainty. Further studies, if necessary, can then focus on predominant parameters. In particular, the constructor can be asked for detailed information only about the most significant parameters.

  6. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  7. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center

    2015-11-15

    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  8. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... intersect. Routes for unicast sessions are agnostic to other sessions and setup beforehand, CORE will then discover and exploit intersecting routes. Our approach allows the inter-session regions to leverage RLNC to compensate for losses or failures in the overhearing or transmitting process. Thus, we...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE...

  9. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... intersect. Routes for unicast sessions are agnostic to other sessions and setup beforehand, CORE will then discover and exploit intersecting routes. Our approach allows the inter-session regions to leverage RLNC to compensate for losses or failures in the overhearing or transmitting process. Thus, we...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE...

  10. Database design for Physical Access Control System for nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Sathishkumar, T., E-mail: satishkumart@igcar.gov.in; Rao, G. Prabhakara, E-mail: prg@igcar.gov.in; Arumugam, P., E-mail: aarmu@igcar.gov.in

    2016-08-15

    Highlights: • Database design needs to be optimized and highly efficient for real time operation. • It requires a many-to-many mapping between Employee table and Doors table. • This mapping typically contain thousands of records and redundant data. • Proposed novel database design reduces the redundancy and provides abstraction. • This design is incorporated with the access control system developed in-house. - Abstract: A (Radio Frequency IDentification) RFID cum Biometric based two level Access Control System (ACS) was designed and developed for providing access to vital areas of nuclear facilities. The system has got both hardware [Access controller] and software components [server application, the database and the web client software]. The database design proposed, enables grouping of the employees based on the hierarchy of the organization and the grouping of the doors based on Access Zones (AZ). This design also illustrates the mapping between the Employee Groups (EG) and AZ. By following this approach in database design, a higher level view can be presented to the system administrator abstracting the inner details of the individual entities and doors. This paper describes the novel approach carried out in designing the database of the ACS.

  11. Impact of the symmetry energy on nuclear pasta phases and crust-core transition in neutron stars

    CERN Document Server

    Bao, S S

    2015-01-01

    We study the impact of the symmetry energy on properties of nuclear pasta phases and crust-core transition in neutron stars. We perform a self-consistent Thomas--Fermi calculation employing the relativistic mean-field model. The properties of pasta phases presented in the inner crust of neutron stars are investigated and the crust-core transition is examined. It is found that the slope of the symmetry energy plays an important role in determining the pasta phase structure and the crust-core transition. The correlation between the symmetry energy slope and the crust-core transition density obtained in the Thomas--Fermi approximation is consistent with that predicted by the liquid-drop model.

  12. Evaluation of nuclear characteristics of DCA modification core for sub-critical measurement

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

    Critical experiments were carried out on Deuterium Critical Assembly (DCA) modification core. DCA modification core has two regions, that is, test region and driver region. The test region consists of various types of fuel and moderator, while the driver region remains the same as the original DCA core (ATR simulated core). Critical characteristics were measured with various types of core patterns and were compared with calculated values based on SCALE code system. Monte Calro code KENO was found to be very accurate in the core analysis. The accuracy stays below 0.5 %dk/k in keff even if core configuration is extremely complicated. (author)

  13. 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...

  14. Modified Anchor Shaped Post Core Design for Primary Anterior Teeth

    OpenAIRE

    R. Rajesh; Kusai Baroudi; K. Bala Kasi Reddy; Praveen, B. H.; V. Sumanth Kumar; Amit, S

    2014-01-01

    Restoring severely damaged primary anterior teeth is challenging to pedodontist. Many materials are tried as a post core but each one of them has its own drawbacks. This a case report describing a technique to restore severely damaged primary anterior teeth with a modified anchor shaped post. This technique is not only simple and inexpensive but also produces better retention.

  15. Modified Anchor Shaped Post Core Design for Primary Anterior Teeth

    Directory of Open Access Journals (Sweden)

    R. Rajesh

    2014-01-01

    Full Text Available Restoring severely damaged primary anterior teeth is challenging to pedodontist. Many materials are tried as a post core but each one of them has its own drawbacks. This a case report describing a technique to restore severely damaged primary anterior teeth with a modified anchor shaped post. This technique is not only simple and inexpensive but also produces better retention.

  16. Modified anchor shaped post core design for primary anterior teeth.

    Science.gov (United States)

    Rajesh, R; Baroudi, Kusai; Reddy, K Bala Kasi; Praveen, B H; Kumar, V Sumanth; Amit, S

    2014-01-01

    Restoring severely damaged primary anterior teeth is challenging to pedodontist. Many materials are tried as a post core but each one of them has its own drawbacks. This a case report describing a technique to restore severely damaged primary anterior teeth with a modified anchor shaped post. This technique is not only simple and inexpensive but also produces better retention.

  17. 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…

  18. 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

  19. Identifying and Using ‘Core Competencies’ to Help Design and Assess Undergraduate Neuroscience Curricula

    Science.gov (United States)

    Kerchner, Michael; Hardwick, Jean C.; Thornton, Janice E.

    2012-01-01

    There has been a growing emphasis on the use of core competencies to design and inform curricula. Based on our Faculty for Undergraduate Neuroscience workshop at Pomona we developed a set of neuroscience core competencies. Following the workshop, faculty members were asked to complete an online survey to determine which core competencies are considered most essential and the results are presented. Backward Design principles are then described and we discuss how core competencies, through a backward design process, can be used to design and assess an undergraduate neuroscience curriculum. Oberlin College is used as a case study to describe the use of core competencies to help develop learning objectives, activities, and assessment measures for an undergraduate neuroscience major. PMID:23494749

  20. Identifying and using 'core competencies' to help design and assess undergraduate neuroscience curricula.

    Science.gov (United States)

    Kerchner, Michael; Hardwick, Jean C; Thornton, Janice E

    2012-01-01

    There has been a growing emphasis on the use of core competencies to design and inform curricula. Based on our Faculty for Undergraduate Neuroscience workshop at Pomona we developed a set of neuroscience core competencies. Following the workshop, faculty members were asked to complete an online survey to determine which core competencies are considered most essential and the results are presented. Backward Design principles are then described and we discuss how core competencies, through a backward design process, can be used to design and assess an undergraduate neuroscience curriculum. Oberlin College is used as a case study to describe the use of core competencies to help develop learning objectives, activities, and assessment measures for an undergraduate neuroscience major.

  1. Uranium Enrichment Reduction in the PGSFR Core Design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chihyung; Hartanto, Donny; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Korea is currently developing the so-called Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) to investigate and demonstrate the capability of TRU transmutation. However, since fuel recycling technology is still at early development in Korea and also due to lack of experience in burning TRU in a fast reactor, the initial core of PGSFR is loaded with low-enriched uranium (LEU) fuel. Several test assemblies containing TRU fuels are supposed to be irradiated and tested for future TRU fuel developments. The uranium enrichment in the LEU PGSFR core is high, about 19.20%, due to large neutron leakage and low conversion ratio. In this paper, the required uranium enrichment is reduced by replacing the reflector material and modifying the reflector geometry in order to decrease the fuel cost of the LEU PGSFR core. PbO is chosen as the reflector material to replace the current HT9 and an inverted reflector assembly is also investigated in this study. It is shown that longer cycle length, higher fuel burnup and flattening power distribution can be achieved with PbO reflector and enhanced neutron leakage can be handled by the optimization of shielding material or core geometry. PbO reflector with inverted geometry is suggest in this research and by using inverted PbO reflector, core performance can be improved while leakage is negligibly enhanced than conventional pin type reflector assembly. Research about reducing the uranium enrichment more by increasing the uranium content in the uranium fuel which is U-10Zr now or increasing the smeared density which is currently 75% can be considered as a future work. Detailed analysis about multi-batch fuel management should be carried out since currently it is done approximately by using linear reactivity theory. Also, analysis for PGSFR with various reflector materials like LME, liquid lead will be carried out and the chemical reaction of those materials including PbO with sodium should be carefully investigated.

  2. Forces in bolted joints: analysis methods and test results utilized for nuclear core applications (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Crescimanno, P.J.; Keller, K.L.

    1981-03-01

    Analytical methods and test data employed in the core design of bolted joints for the LWBR core are presented. The effects of external working loads, thermal expansion, and material stress relaxation are considered in the formulation developed to analyze joint performance. Extensions of these methods are also provided for bolted joints having both axial and bending flexibilities, and for the effect of plastic deformation on internal forces developed in a bolted joint. Design applications are illustrated by examples.

  3. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-05-15

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  4. Design of transition cores of RSG GAS (MPR-30) with higher loading silicide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Liem, Peng Hong, E-mail: liemph@nais.ne.j [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro, Tokyo 152-8550 (Japan); Sembiring, Tagor Malem [Center for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency (Batan), Puspiptek, Serpong, Tangerang 15310 (Indonesia)

    2010-06-15

    A procedure of designing transition cores to achieve the equilibrium silicide core of RSG GAS with higher fuel loading of 300 g U/fuel element (FE) (meat density of 3.55 g U/cm{sup 3}) has been proposed. In the proposed procedure, the EOC excess reactivity of each transition core is minimized in order to satisfy the safety design limit of one-stuck-rod sub-criticality margin while keeping the maximum of radial power peaking factor below the allowable value. Under the design procedure, the initial fuel loadings are increased gradually in two steps, i.e. from 250 to 275 g U/FE followed by 275-300 g U/FE. The analysis results show that all transition cores can satisfy all design requirements and safety limits. We concluded that the obtained transition core design should be adopted into the future core conversion program of RSG GAS. The targeted silicide core can be achieved practically in at least 24 transition cores.

  5. Future CANDU nuclear power plant design requirements document executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Usmani, S.A. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new.

  6. Designing tools for oil exploration using nuclear modeling

    Directory of Open Access Journals (Sweden)

    Mauborgne Marie-Laure

    2017-01-01

    Full Text Available When designing nuclear tools for oil exploration, one of the first steps is typically nuclear modeling for concept evaluation and initial characterization. Having an accurate model, including the availability of accurate cross sections, is essential to reduce or avoid time consuming and costly design iterations. During tool response characterization, modeling is benchmarked with experimental data and then used to complement and to expand the database to make it more detailed and inclusive of more measurement environments which are difficult or impossible to reproduce in the laboratory. We present comparisons of our modeling results obtained using the ENDF/B-VI and ENDF/B-VII cross section data bases, focusing on the response to a few elements found in the tool, borehole and subsurface formation. For neutron-induced inelastic and capture gamma ray spectroscopy, major obstacles may be caused by missing or inaccurate cross sections for essential materials. We show examples of the benchmarking of modeling results against experimental data obtained during tool characterization and discuss observed discrepancies.

  7. Software Design Document for the AMP Nuclear Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Philip, Bobby [ORNL; Clarno, Kevin T [ORNL; Cochran, Bill [ORNL

    2010-03-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  8. 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

  9. Design principles of a nuclear and industrial HVAC of IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Pruneri, Giuseppe [IFMIF/EVEDA, Project Team, Rokkasho (Japan); Ibarra, A. [CIEMAT, Madrid (Spain); Heidinger, R. [F4E, Garching (Germany); Knaster, J. [IFMIF/EVEDA Project Team, Rokkasho (Japan); Sugimoto, M. [JAEA, Rokkasho (Japan)

    2016-02-15

    Highlights: • Parameter of Derivate air Contamination (DAC) allows to associate the type of air ventilation. • The construction and operation of IFMIF will be subjected to the regulations of the country in which it will be sited. • Structures, systems and components are assigned a particular safety important components (SIC, 1–2 and Non-SIC) clarification that is based on the consequences of their failure. • Reliability, Availability, Maintainability and Inspectability (RAMI) analysis has given a great contribution of the facility to optimize the configuration, particularly for the HVAC system. - Abstract: In 2013, the IFMIF, the International Fusion Material Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase, framed by the Broader Approach Agreement between Japan and EURATOM, accomplished in 2013 its mandate to provide the engineering design of the plant on schedule [1]. The IFMIF aims to qualify and characterize materials that are capable of withstanding the intense neutron flux originated in D-T reactions of future fusion reactors due to a neutron flux with a broad peak at 14 MeV, which is able to provide >20 dpa/fpy on small specimens in this EVEDA phase. The successful operation of such a challenging plant demands a careful assessment of the Conventional Facilities (CF), which have adequate redundancies to allow for the target plant availability [2]. The present paper addresses the design proposed in the IFMIF Intermediate Engineering Design Report regarding the CF, particularly the IFMIF's Nuclear and Industrial HVAC design. A preliminary feasibility study, including the initial configuration, calculations and reliability/availability analysis, were performed. The nuclear HVAC design was developed progressively; first, by establishing a conceptual design, starting from the system functional description, followed by the identification of the corresponding interfacing systems and their

  10. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Science.gov (United States)

    Schnitzler, Bruce G.; Borowski, Stanley K.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were

  11. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine

  12. Polymer Design and Processing for Liquid-Core waveguides

    DEFF Research Database (Denmark)

    Sagar, Kaushal Shashikant

    photochemistry via UV photo-oxidation of nanoporous 1,2-PB. Detailed quantitative and qualitative analysis of photo-oxidation in the presence of air is carried out by gravimetry, titrimetry and spectrometry. Distribution study of the hydrophilic photo-products relative to the polymer-air interface shows high...... of the photo grafting reaction on the nanoporous wall are studied using gravimetry. The fabrication of solid-liquid core waveguides is done by adapting the know-how on thiol-ene photochemistry to standard microfabrication cleanroom setup and UV lithography. Contrast curves for thiol-ene systems are reported...

  13. Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

    Science.gov (United States)

    Shirvan, Koroush

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

  14. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Mexico, D.F. (Mexico); Francois, Juan Luis [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico)]. E-mail: jlfl@fi-b.unam.mx; Martin-del-Campo, Cecilia [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana, Avenida San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico)

    2005-04-15

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the {sup 233}U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly.

  15. IPE Data Base: Plant design, core damage frequency and containment performance information

    Energy Technology Data Exchange (ETDEWEB)

    Lehner, J.; Lin, C.C.; Pratt, W.T. [Brookhaven National Lab., Upton, NY (United States); Su, T.; Danziger, L. [Nuclear Regulatory Commission, Rockville, MD (United States)

    1995-12-31

    This data base stores data obtained from the Individual Plant Examinations (IPEs) which licensees of nuclear power plants have conducted in response to NRC`s Generic Letter GL88-20. The IPE Data Base is a collection of linked files which store information about plant design, core damage frequency, and containment performance in a uniform, structured way. The information contined in the various files is based on data contained in the IPE submittals. The information extracted from the submittals and entered into the IPE Data Base can be maniulated so that queries regarding individual or groups of plants can be answered using the IPE Data Base. The IPE Data Base supports detailed inquiries into the characteristics of individual plants or classes of plants. Progress has been made on the IPE Data Base and it is largely complete. Recent focus has been the development of a user friendly version which is menu driven and allows the user to ask queries of varying complexity easily, without the need to become familiar with particular data base formats or conventions such as those of DBase IV or Microsoft Access. The user can obtain the information he desired by quickly moving through a series of on-screen menus and ``clicking`` on appropriate choices. In this way even a first time user can benefit from the large amount of information stored in the IPE Data Base without the need of a learning period.

  16. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (Dcore, when we use the balance geometry, the k-eff value flattest and more stable than the others.

  17. Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. E-mail: robbe@aquilon.cea.frmfrobbe@cea.fr; Lepareux, M.; Treille, E.; Cariou, Y

    2003-08-01

    In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant. The MARA 10 experimental test simulates a HCDA in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge. This paper presents a numerical simulation of the test with the EUROPLEXUS code and an analysis of the computed results. In particular, the evolution of the fluid flows and the deformations of the internal and external structures are analysed in detail. Finally, the current computed results are compared with the experimental ones and with previous numerical results computed with the SIRIUS and CASTEM-PLEXUS codes.

  18. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    Energy Technology Data Exchange (ETDEWEB)

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  19. Seismic design and analysis of nuclear power plant structures

    Institute of Scientific and Technical Information of China (English)

    Pentti Varpasuo

    2013-01-01

    The seismic design and analysis of nuclear power plant (NPP) begin with the seismic hazard assessment and design ground motion development for the site.The following steps are needed for the seismic hazard assessment and design ground motion development:a.the development of regional seismo-tectonic model with seismic source areas within 500 km radius centered to the site; b.the development of strong motion prediction equations;c.logic three development for taking into account uncertainties and seismic hazard quantification; d.the development of uniform hazard response spectra for ground motion at the site; e.simulation of acceleration time histories compatible with uniform hazard response spectra.The following phase two in seismic design of NPP structures is the analysis of structural response for the design ground motion.This second phase of the process consists of the following steps:a.development of structural models of the plant buildings; b.development of the soil model underneath the plant buildings for soil-structure interaction response analysis; c.determination of in-structure response spectra for the plant buildings for the equipment response analysis.In the third phase of the seismic design and analysis the equipment is analyzed on the basis of in-structure response spectra.For this purpose the structural models of the mechanical components and piping in the plant are set up.In large 3D-structural models used today the heaviest equipment of the primary coolant circuit is included in the structural model of the reactor building.In the fourth phase the electrical equipment and automation and control equipment are seismically qualified with the aid of the in-structure spectra developed in the phase two using large three-axial shaking tables.For this purpose the smoothed envelope spectra for calculated in-structure spectra are constructed and acceleration time is fitted to these smoothed envelope spectra.

  20. 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.

  1. Designing, Leading and Managing the Transition to the Common Core: A Strategy Guidebook for Leaders

    Science.gov (United States)

    Brown, Brentt; Vargo, Merrill

    2014-01-01

    The Common Core provides districts an opportunity to renew their focus on teaching and learning. But it also poses a number of design and implementation challenges for school districts. The "Leadership and Design Cycles" described in this guidebook offers an evidenced-based and structured process for leaders to design and implement…

  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. Optical design of a laser system for nuclear fusion research.

    Science.gov (United States)

    de Metz, J

    1971-07-01

    High power laser improvements, high quality aspheric lenses, and sharp focusing on a solid deuterium target enable us to get numerous nuclear fusion reactions inside the deuterium plasma. Since Maiman successfully built the first light amplifier in 1960 [Nature 187, 493 (1960)] and Terhune performed air breakdown experiments in 1962 ["Optical Third Harmonic Generation," Comptes rendus de la 3ème Conférence Internationale d'Electronique Quantique, Paris, 11-15 février 1963, P. Grivet and N. Bloembergen, Eds. (Dunod, Paris, 1964), pp. 1559-15761, the laser has been thought of as a valuable energy source for fusion devices. Now a kind of race has started toward high temperature plasmas created by powerful lasers. However, the peak power of solid state laser is limited by glass damage, pump efficiences, and unwanted effects such as superradiance. So it is necessary to improve all the optical properties of the laser and the focusing of the lens on the target. In this paper, requirements for fusion implying a very high flux will be stated. Successive optical designs will be described together with measurement methods, and the contribution of optical improvements to the occurrence of nuclear fusion reaction in deuterium targets will be evaluated.

  4. An Assessment of Testing Requirement Impacts on Nuclear Thermal Propulsion Ground Test Facility Design

    Science.gov (United States)

    Shipers, Larry R.; Ottinger, Cathy A.; Sanchez, Lawrence C.

    1994-07-01

    Programs to develop solid core nuclear thermal propulsion (NTP) systems have been under way at the Department of Defense (DoD), the National Aeronautics and Space Administration (NASA), and the Department of Energy (DOE). These programs have recognized the need for a new ground test facility to support development of NTP systems. However, the different military and civilian applications have led to different ground test facility requirements. The Department of Energy (DOE) in its role as landlord and operator of the proposed research reactor test facilities has initiated an effort to explore opportunities for a common ground test facility to meet both DoD and NASA needs. The baseline design and operating limits of the proposed DoD NTP ground test facility are described. The NASA ground test facility requirements are reviewed and their potential impact on the DoD facility baseline is discussed.

  5. An assessment of testing requirement impacts on nuclear thermal propulsion ground test facility design

    Energy Technology Data Exchange (ETDEWEB)

    Shipers, L.R.; Ottinger, C.A.; Sanchez, L.C.

    1993-10-25

    Programs to develop solid core nuclear thermal propulsion (NTP) systems have been under way at the Department of Defense (DoD), the National Aeronautics and Space Administration (NASA), and the Department of Energy (DOE). These programs have recognized the need for a new ground test facility to support development of NTP systems. However, the different military and civilian applications have led to different ground test facility requirements. The Department of Energy (DOE) in its role as landlord and operator of the proposed research reactor test facilities has initiated an effort to explore opportunities for a common ground test facility to meet both DoD and NASA needs. The baseline design and operating limits of the proposed DoD NTP ground test facility are described. The NASA ground test facility requirements are reviewed and their potential impact on the DoD facility baseline is discussed.

  6. Verification and uncertainty evaluation of HELIOS/MASTER nuclear design system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jae Seung; Kim, J. C.; Cho, B. O. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    A nuclear design system HELIOS/MASTER was established and core follow calculations were performed for Yonggwang Unit 1 cycles 1 through 7 and Yonggwang Unit 3 cycles 1 through 2. The accuracy of HELIOS/MASTER system was evaluated by estimations of uncertainties of reactivity and peaking factors and by comparisons of the maximum differences of isothermal temperature coefficient, inverse boron worth and control rod worth with the CASMO-3/MASTER uncertainties. The reactivity uncertainty was estimated by 362 pcm, and the uncertainties of three-dimensional, axially integrated radial, and planar peaking factors were evaluated by 0.048, 0.034, and 0.044 in relative power unit, respectively. The maximum differences of isothermal temperature coefficient, inverse boron worth and control rod worth were within the CASMO-3/MASTER uncertainties. 17 refs., 17 figs., 10 tabs. (Author)

  7. Design of low-loss and highly birefringent hollow-core photonic crystal fiber

    Science.gov (United States)

    Roberts, P. J.; Williams, D. P.; Sabert, H.; Mangan, B. J.; Bird, D. M.; Birks, T. A.; Knight, J. C.; Russell, P. St. J.

    2006-08-01

    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 on the core-surround ring, gives rise to wavelength ranges where the effective index difference between the polarization modes is larger than 10-4. At such high birefringence levels, one of the polarization modes retains favorable field exclusion characteristics, thus enabling low-loss propagation of this polarization channel.

  8. Design of large-core single-mode Yb3+-doped photonic crystal fiber

    Institute of Scientific and Technical Information of China (English)

    ZHAO Xing-tao; ZHENG Yi; LIU Xiao-xu; ZHOU Gui-yao; LIU Zhao-lun; HOU Lan-tian

    2012-01-01

    The effective index of the cladding fundamental space-filing mode in photonic crystal fiber (PCF) is simulated by the effective index method.The variation of the effective index with the structure parameters of the fiber is achieved.For thefirst thne,the relations of the V parameter ofYb3+-doped PCF with the refractive index of core and the structure parameters of the fiber are provided.The single-mode characteristics of large-core yb3+-doped photonic crystal fibers with 7 and 19 missing air holes in the core are analyzed.The large-core single-mode Yb3+-doped photonic crystal fibers with core diameters of 50 μm,100 μm and 150 μm are designed.The results provide theory instruction for the design and fabrication of fiber.

  9. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    Energy Technology Data Exchange (ETDEWEB)

    Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

    2015-06-15

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

  10. 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...

  11. Melt spreading code assessment, modifications, and application to the EPR core catcher design.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T .; Nuclear Engineering Division

    2009-03-30

    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

  12. Improvements of seismic design of nuclear power plant equipment

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Kohei [Tokyo Metropolitan Univ., Hachioji (Japan). Faculty of Technology; Takayama, Yoshihiro

    1997-03-01

    A brief survey and overview of the current research and development in Japan was presented. Particularly, several kinds of new dampers and isolators were developed and those effectiveness were examined by caring out the large-scale vibration test and so on. The evaluation of the energy absorption of these damping devices at the earthquake appeared to be significant. In addition, it must be necessary to investigate the design margin and the failure mode and limit problem to these devices and the nuclear structures and piping supported by those. Mutual exchange of the information related to these technology and research has to be put forward and cooperative works including the international conference on those issues should be promoted. (J.P.N.)

  13. Parameter Study of the LIFE Engine Nuclear Design

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Meier, W R; Latkowski, J F; Abbott, R P

    2009-07-10

    LLNL is developing the nuclear fusion based Laser Inertial Fusion Energy (LIFE) power plant concept. The baseline design uses a depleted uranium (DU) fission fuel blanket with a flowing molten salt coolant (flibe) that also breeds the tritium needed to sustain the fusion energy source. Indirect drive targets, similar to those that will be demonstrated on the National Ignition Facility (NIF), are ignited at {approx}13 Hz providing a 500 MW fusion source. The DU is in the form of a uranium oxycarbide kernel in modified TRISO-like fuel particles distributed in a carbon matrix forming 2-cm-diameter pebbles. The thermal power is held at 2000 MW by continuously varying the 6Li enrichment in the coolants. There are many options to be considered in the engine design including target yield, U-to-C ratio in the fuel, fission blanket thickness, etc. Here we report results of design variations and compare them in terms of various figures of merit such as time to reach a desired burnup, full-power years of operation, time and maximum burnup at power ramp down and the overall balance of plant utilization.

  14. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  15. Axial design of nuclear fuel using path relinking; Diseno axial de combustible nuclear utilizando path relinking

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, A.; Torres, M.; Ortiz, J. J.; Perusquia, R.; Hernandez, J. L.; Montes, J. L. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jacm@nuclear.inin.mx

    2008-07-01

    In the present work the preliminary results were obtained with the zoctli system whose purpose is the axial design of assembly of nuclear fuel under certain considerations. For the mentioned design well-know cells were already used and that they have been proven in diverse cycles of operation in the nuclear power plant of Laguna Verde. The design contemplates fuels assemblies of 10x10 and with 2 water channels. The assembly was distributed in 6 axial zones according to its structure. In order to take to end the optimization is was used the well-known technique like Path relinking and to find the group of previous solutions required by this technique uses the technical Taboo search. In order to work with Path relinking, 5 trajectories was taken in to account from a set of 5 previous solutions generated with theTaboo search, the update of the group of solutions is carried out in dynamic form. In the case of the Taboo search it was used a list of variable size, it was implement an aspiration approach, it was used the vector of frequencies and due to the cost of the evaluation of the objective function, only it was review 5% of the vicinity. For the objective function was considered the limit thermal, the axial profile of power, the effective multiplication factor and the margin of having turned off in cold. In order to prove the design system, it was used a balance cycle with a value of reference of 0.9928 for the effective multiplication factor that is equivalent to a produced energy of 10896 MWd/TU at the end of operation to full power. The designed assemblies were placed both in one of lots different from fresh assemblies on which it counts the referred cycle. At the end one a comparison with the results obtained with other techniques and under similar conditions is made. The results obtained until the moment show an appropriate performance of the system. It is possible to indicate that a small inconvenient is the amount of consumed resources of calculation during

  16. Materials for Nuclear Plants From Safe Design to Residual Life Assessments

    CERN Document Server

    Hoffelner, Wolfgang

    2013-01-01

    The clamor for non-carbon dioxide emitting energy production has directly  impacted on the development of nuclear energy. As new nuclear plants are built, plans and designs are continually being developed to manage the range of challenging requirement and problems that nuclear plants face especially when managing the greatly increased operating temperatures, irradiation doses and extended design life spans. Materials for Nuclear Plants: From Safe Design to Residual Life Assessments  provides a comprehensive treatment of the structural materials for nuclear power plants with emphasis on advanced design concepts.   Materials for Nuclear Plants: From Safe Design to Residual Life Assessments approaches structural materials with a systemic approach. Important components and materials currently in use as well as those which can be considered in future designs are detailed, whilst the damage mechanisms responsible for plant ageing are discussed and explained. Methodologies for materials characterization, material...

  17. Safeguards Guidance Document for Designers of Commercial Nuclear Facilities: International Nuclear Safeguards Requirements and Practices For Uranium Enrichment Plants

    Energy Technology Data Exchange (ETDEWEB)

    Robert Bean; Casey Durst

    2009-10-01

    This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called “Safeguards-by-Design.” This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, “International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials,” published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a

  18. Aseismic Design Licensings and guidelines for nuclear power plant in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yoshizawa, Kazumi [Agency of Natural Resources and Energy, Tokyo (Japan)

    1997-03-01

    This paper describes Aseismic Design Licensing for Japanese Nuclear Power Plants which includes system, procedures and brief contents concerned application, permit and inspection, and the `Examination Guide for Aseismic Design of the Nuclear Power Reactor Facilities` which focused principals of seismic design loads, load combinations, and allowable limits. (J.P.N.)

  19. Study of seismic design bases and site conditions for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-01

    This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches.

  20. MOX燃料堆芯热工特性及设计限值研究%Analysis of MOX core thermal characteristics and design criteria

    Institute of Scientific and Technical Information of China (English)

    刘一哲; 喻宏; 田和春

    2009-01-01

    For the MOX core of a Sodium-cooled Fast Reactor(SFR) nuclear power plant, the advantages are higher linear power, burn-up and outlet temperature. Core thermal hydraulic design meets some new issues. Based on the analysis of MOX fuel characteristics, the thermal design criteria of core were raised in this paper. Core thermal characteristics and margin were analyzed for the 870 MWe nuclear power plant. The results showed that higher thermal parameters were reasonable and feasible for the MOX core, and the thermal margin satisfied the requirements.%使用MOX燃料的快堆核电站以其线功率高、燃耗高、堆芯出口温度高等特点,对堆芯热工设计提出了新的问题.本文在对MOX燃料热工性能分析的基础上,给出了主要的热工设计限值,并以电功率870 MW电站为参考,初步分析了其堆芯热工特性和设计裕量.结果表明对于MOX燃料,较高的堆芯热工参数合理可行,且具有足够的裕量.

  1. Modified Y-TZP core design improves all-ceramic crown reliability.

    Science.gov (United States)

    Silva, N R F A; Bonfante, E A; Rafferty, B T; Zavanelli, R A; Rekow, E D; Thompson, V P; Coelho, P G

    2011-01-01

    This study tested the hypothesis that all-ceramic core-veneer system crown reliability is improved by modification of the core design. We modeled a tooth preparation by reducing the height of proximal walls by 1.5 mm and the occlusal surface by 2.0 mm. The CAD-based tooth preparation was replicated and positioned in a dental articulator for core and veneer fabrication. Standard (0.5 mm uniform thickness) and modified (2.5 mm height lingual and proximal cervical areas) core designs were produced, followed by the application of veneer porcelain for a total thickness of 1.5 mm. The crowns were cemented to 30-day-aged composite dies and were either single-load-to-failure or step-stress-accelerated fatigue-tested. Use of level probability plots showed significantly higher reliability for the modified core design group. The fatigue fracture modes were veneer chipping not exposing the core for the standard group, and exposing the veneer core interface for the modified group.

  2. Design analysis and risk assessment for a single stage to orbit nuclear thermal rocket

    Science.gov (United States)

    Labib, Satira I.

    Recent advances in high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This thesis describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1-15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 700 seconds. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. At the same power level, the 40 cm reactor results in the lowest radiation dose rate of the three reactors. Radiation dose rates decrease to background levels ~3.5 km from the launch site. After a one-year decay time, all of the activated materials produced by an NTR launch would be classified as Class A low-level waste. The activation of air produces significant amounts of argon-41 and nitrogen-16 within 100 m of the launch. The derived air concentration, DAC, from the activation products decays to less than unity within two days, with only argon-41 remaining. After 10 minutes of full power operation the 120 cm core corresponding to a 15 MT payload contains 2.5 x 1013, 1.4 x 1012, 1.5 x 1012, and 7.8 x 10 7 Bq of 131I, 137Cs, 90Sr, and 239Pu respectively. The decay heat after shutdown increases with increasing reactor power with a maximum decay heat of 108 kW immediately after shutdown for the 15 MT payload.

  3. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  4. Human factors design guidelines for maintainability of Department of Energy nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bongarra, J.P. Jr.; VanCott, H.P.; Pain, R.F.; Peterson, L.R.; Wallace, R.I.

    1985-06-18

    Intent of these guidelines is to provide design and design review teams of DOE nuclear facilities with human factors principles to enhance the design and aid in the inspection of DOE nuclear facilities, systems, and equipment. These guidelines are concerned with design features of DOE nuclear facilities which can potentially affect preventive and corrective maintenance of systems within DOE nuclear facilities. Maintenance includes inspecting, checking, troubleshooting, adjusting, replacing, repairing, and servicing activities. Other factors which influence maintainability such as repair and maintenance suport facilities, maintenance information, and various aspects of the environment are also addressed.

  5. Design and Testing of an Active Core for Sandwich Panels

    Science.gov (United States)

    2008-03-01

    structures such as the Kagome truss (Hutchinson, Wicks et al. 2003; Symons, Hutchinson et al. 2005) holds potential in the design of morphing structures that...New Yourk, John Wiley and Sons, Inc. Hutchinson, R. G., N. Wicks, et al. (2003). " Kagome plate structures for actuation." International Journal of...Optimization 12(2): 18. Symons, D. D., R. G. Hutchinson, et al. (2005). "Actuation of the Kagome double layer grid part 1: Prediction of performance

  6. Design issues on using FPGA-based I and C systems in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Santos, Isaac Jose A.L. dos; Lacerda, Fabio de, E-mail: msantana@ien.gov.br, E-mail: paulov@ien.gov.br, E-mail: luquetti@ien.gov.br, E-mail: acerda@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Div. de Engenharia Nuclear

    2015-07-01

    The FPGA (field programmable gate array) is widely used in various fields of industry. FPGAs can be used to perform functions that are safety critical and require high reliability, like in automobiles, aircraft control and assistance and mission-critical applications in the aerospace industry. With these merits, FPGAs are receiving increased attention worldwide for application in nuclear plant instrumentation and control (I and C) systems, mainly for Reactor Protection System (RPS). Reasons for this include the fact that conventional analog electronics technologies are become obsolete. I and C systems of new Reactors have been designed to adopt the digital equipment such as PLC (Programmable Logic Controller) and DCS (Distributed Control System). But microprocessors-based systems may not be simply qualified because of its complex characteristics. For example, microprocessor cores execute one instruction at a time, and an operating system is needed to manage the execution of programs. In turn, FPGAs can run without an operating system and the design architecture is inherently parallel. In this paper we aim to assess these and other advantages, and the limitations, on FPGA-based solutions, considering the design guidelines and regulations on the use of FPGAs in Nuclear Plant I and C Systems. We will also examine some circuit design techniques in FPGA to help mitigate failures and provide redundancy. The objective is to show how FPGA-based systems can provide cost-effective options for I and C systems in modernization projects and to the RMB (Brazilian Multipurpose Reactor), ensuring safe and reliable operation, meeting licensing requirements, such as separation, redundancy and diversity. (author)

  7. Nuclear heating measurements by in-pile calorimetry: prospective works for a microsensor design

    Energy Technology Data Exchange (ETDEWEB)

    Reynard-Carette, C.; Carette, M.; Aguir, K.; Bendahan, M.; Fiorido, T. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A.; Fourmentel, D.; Villard, J.F. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 (France); Barthes, M.; Lanzetta, F.; Layes, G.; Vives, S. [FEMTO-ST, UMR 6174, Departement ENERGIE, Universite de Franche-Comte, 90000, Belfort (France)

    2015-07-01

    Since 2009 works have been performed in the framework of joint research programs between CEA and Aix-Marseille University. The main aim of these programs is to design and develop in-pile instrumentations, advanced calibration procedure and accurate measurement methods in particular for the new Material Testing Reactor (MTR) under construction in the South of France: Jules Horowitz Reactor (JHR). One major sensor is a specific radiometric calorimeter, which was studied out-of-pile from a thermal point of view and in-pile during irradiation campaigns. This sensor type is dedicated to measurements of nuclear heating (energy deposition rate per mass unit induced by interactions between nuclear rays and matter) inside experimental channels of MTRs. This kind of in-pile calorimeter corresponds to heat flux calorimeter exchanging with the external cooling fluid. This thermal running mode allows the establishment of steady thermal conditions inside the sensor to carry out online continuous measurements inside the reactor (core or reflector). Two main types of calorimeters exist. The first type consists of a single cell calorimeter. It is divided into a sample of material to be tested and a jacket instrumented with two thermocouples or a single thermocouple (Gamma Thermometer). The second, called a differential calorimeter, is composed of two superposed twin cells (a measurement cell containing a sample of material, and a reference cell to remove the heating of the cell body) instrumented with four thermocouples and two electrical heaters. Contrary to a single-cell calorimeter, a differential calorimeter allows the compensation of the parasite nuclear heating of the sensor body or jacket. Moreover, it possesses interesting advantages: thanks to the heaters embedded in the cells, three different measurement methods can be applied during irradiations to quantify nuclear heating. The first one is based on the use of out-of-pile calibration curves obtained by generating a heat

  8. First in-core simultaneous measurements of nuclear heating and thermal neutron flux obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lepeltier, Valerie; Bubendorff, Jacques; Carcreff, Hubert [Nuclear studies and reactor irradiation Service, CEA Saclay 91191 Gif sur Yvette (France); Salmon, Laurent [Thermalhydraulics and Fluid Mechanics Section, CEA Saclay 91191 Gif sur Yvette, (France)

    2015-07-01

    Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heating rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 10{sup 14} n.cm{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by

  9. Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study

    Energy Technology Data Exchange (ETDEWEB)

    CLEVELAND, K.J.

    2000-08-17

    This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage.

  10. 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...

  11. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  12. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). About cores, coal and cash

    Energy Technology Data Exchange (ETDEWEB)

    Podivinsky, Tomas Jan

    2016-07-15

    Rationality and - especially with regard to reducing emissions - technological neutrality are two commitments for nuclear fission. The Czech Republic, where conditions are not suitable for economical large-scale operation of facilities based on renewables, there is no alternative in environmental or business policy to the reasonable use of nuclear energy. The aim of the updated Czech energy strategy is to increase the proportion of nuclear energy from 35 % to approx. 50 % of power generation and to cover the rest - together with ultra-high efficiency coal fired power plants - with energy from renewable sources and gas fired power plants.

  13. Challenges in spent nuclear fuel final disposal:conceptual design models

    Institute of Scientific and Technical Information of China (English)

    Mukhtar Ahmed RANA

    2008-01-01

    The disposal of spent nuclear fuel is a long-standing issue in nuclear technology. Mainly, UO2 and metallic U are used as a fuel in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements, which would remain radioactive for 104 to 108 years. In this brief communication, essential concepts and engineering elements related to high-level nuclear waste disposal are described. Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste. Notions of physical and chemical barriers to contain nuclear waste are highlightened. Concerns regarding integrity, self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed. The question of retrievability of spent nuclear fuel after disposal is considered.

  14. An integrated design methodology for the safety and security of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Joung, S. Y.; Chang, S. H. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2012-03-15

    After Fukushima nuclear power plant accident, safety of nuclear power plant was issued. In Fukushima accident, one of main reason is location of emergency diesel power generators stop which locate under the sea water level when tsunami occurred. In view of security, emergency diesel generator location under reactor building design is good because for example to escape air strike but is not good for safety for example Fukushima accident. Sometimes safety and security design looks conflicting but nuclear safety and nuclear security share the goal of protecting but nuclear safety and nuclear power plants operate at acceptable risk levels. The purpose of this paper is to introduce safety and security integrated design for nuclear power plant in special emergency diesel generator and control room with simple probabilistic safety assessment analysis.

  15. Preliminary issues associated with the next generation nuclear plant intermediate heat exchanger design

    Science.gov (United States)

    Natesan, K.; Moisseytsev, A.; Majumdar, S.

    2009-07-01

    The Next Generation Nuclear Plant, with emphasis on production of both electricity and hydrogen, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 850-950 °C. In this concept, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, a nitrogen/helium mixture, or a molten salt. This paper assesses the issues pertaining to shell-and-tube and compact heat exchangers. A detailed thermal-hydraulic analysis was performed to calculate heat transfer, temperature distribution, and pressure drop inside both printed circuit and shell-and-tube heat exchangers. The analysis included evaluation of the role of key process parameters, geometrical factors in heat exchanger designs, and material properties of structural alloys. Calculations were performed for helium-to-helium, helium-to-helium/nitrogen, and helium-to-salt heat exchangers.

  16. Improved damage tolerant face/core interface design in sandwich structures

    DEFF Research Database (Denmark)

    Lundsgaard-Larsen, Christian; Berggreen, Christian; Quispitupa, Amilcar

    2009-01-01

    kinking behavior may be altered / avoided by changing the interface design by using Chopped Strand Mat (CSM), Continuous Filament Mat (CFM) and woven mats at the face/core interface as sources for fiber bridging, thus keeping and arresting the crack in the interface.......A face/core debond in a sandwich structure may propagate in the interface or kink into either the face or core depending on the mode-mixity of the loading. This study explores experimental methodologies for mapping the kinking behavior at various mode-mixities. Further, it is shown that the crack...

  17. Efficient optimization of hollow-core photonic crystal fiber design using the finite-element method

    DEFF Research Database (Denmark)

    Holzlöhner, Ronald; Burger, Sven; Roberts, John;

    2006-01-01

    We employ a finite-element (FE) solver with adaptive grid refinement to model hollow-core photonic crystal fibers (HC-PCFs) whose core is formed from 19 omitted cladding unit cells. We optimize the complete fiber geometry for minimal field intensity at material/air interfaces, which indicates low...... loss and high damage threshold, using multidimensional optimization. The optimal design shows a 99.8 % power fraction within the air and an overlap with a Gaussian mode of 96.9 %....

  18. Dynamical analysis of innovative core designs facing unprotected transients with the MAT5 DYN code

    Energy Technology Data Exchange (ETDEWEB)

    Darmet, G.; Massara, S. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Since 2007, advanced Sodium-cooled Fast Reactors (SFR) are investigated by CEA, AREVA and EDF in the framework of a joint French collaboration. A prototype called ASTRID, sets out to demonstrate progress made in SFR technology, is due to operate in the years 2020's. The modeling of unprotected transients by computer codes is one of the key safety issues in the design approach to such SFR systems. For that purpose, the activity on CATHARE, which is the reference code for the transient analysis of ASTRID, has been strengthened during last years by CEA. In the meantime, EDF has developed a simplified and multi-channel code, named MAT5 DYN, to analyze and validate innovative core designs facing protected and unprotected transients. First, the paper consists in a description of MAT5 DYN: a code based on the existing code MAT4 DYN including major improvements on geometry description and physical modeling. Second, two core designs based on the CFV core design developed at CEA are presented. Then, the dynamic response of those heterogeneous cores is analyzed during unprotected loss of flow (ULOF) transient and unprotected transient of power (UTOP). The results highlight the importance of the low void core effect specific to the CFV design. Such an effect, when combined with a sufficient primary pump halving time and an optimized cooling group scheme, allows to delay (or, possibly, avoid) the sodium boiling onset during ULOF accidents. (authors)

  19. The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

    2017-01-01

    The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

  20. Single hepatitis-B virus core capsid binding to individual nuclear pore complexes in Hela cells.

    Science.gov (United States)

    Lill, Yoriko; Lill, Markus A; Fahrenkrog, Birthe; Schwarz-Herion, Kyrill; Paulillo, Sara; Aebi, Ueli; Hecht, Bert

    2006-10-15

    We investigate the interaction of hepatitis B virus capsids lacking a nuclear localization signal with nuclear pore complexes (NPCs) in permeabilized HeLa cells. Confocal and wide-field optical images of the nuclear envelope show well-spaced individual NPCs. Specific interactions of capsids with single NPCs are characterized by extended residence times of capsids in the focal volume which are characterized by fluorescence correlation spectroscopy. In addition, single-capsid-tracking experiments using fast wide-field fluorescence microscopy at 50 frames/s allow us to directly observe specific binding via a dual-color colocalization of capsids and NPCs. We find that binding occurs with high probability on the nuclear-pore ring moiety, at 44 +/- 9 nm radial distance from the central axis.

  1. Study on core concept for commercial LMFBR plant toward self-consistent nuclear energy system concept

    Energy Technology Data Exchange (ETDEWEB)

    Toukura, A. [Institute of Applied Energy, Tokyo (Japan); Yamazaki, M. [Toshiba Corp., Fuchu, Tokyo (Japan). Fuchu Works; Ohashi, M. [Hitachi Ltd., Ibaraki (Japan). Hitachi Works; Ikeda, K. [Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan); Saito, M.; Fujiie, Y. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1995-12-31

    Fast Breeder Reactor (FBR) is expected to be commercialized in Japan to overcome foreseeable problems such as reactor safety, increasing energy demand, final disposal of high level radioactive waste and fuel resource shortage. We have been studying three FBR core concepts enhancing its potential abilities; ultra-large type, simplified type and friendly to fuel cycle type core. This study is sponsored by Ministry of International Trade and Industry. (author).

  2. Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

    2015-10-15

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

  3. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  4. Health physics activities in support of the thermal shield removal/disposal and core support barrel repair at the St. Lucie Nuclear Power Plant.

    Science.gov (United States)

    Maisler, J J; Buchanan, H F

    1988-02-01

    The health physics activities related to the removal and disposal of a thermal shield at a nuclear power plant and subsequent repairs to the core support barrel required increased planning relative to a normal refueling/maintenance outage. The repair of the core support barrel was a "first" in the nuclear power industry. Pre-job planning was of great concern because of extremely high radiation levels associated with the irradiated stainless steel thermal shield and core support barrel. ALARA techniques used in the preparation of the thermal shield for removal and shipment to the disposal site are discussed.

  5. Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

  6. Conceptual design report: Nuclear materials storage facility renovation. Part 1, Design concept. Part 2, Project management

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-14

    The Nuclear Materials Storage Facility (NMSF) at the Los Alamos National Laboratory (LANL) was a Fiscal Year (FY) 1984 line-item project completed in 1987 that has never been operated because of major design and construction deficiencies. This renovation project, which will correct those deficiencies and allow operation of the facility, is proposed as an FY 97 line item. The mission of the project is to provide centralized intermediate and long-term storage of special nuclear materials (SNM) associated with defined LANL programmatic missions and to establish a centralized SNM shipping and receiving location for Technical Area (TA)-55 at LANL. Based on current projections, existing storage space for SNM at other locations at LANL will be loaded to capacity by approximately 2002. This will adversely affect LANUs ability to meet its mission requirements in the future. The affected missions include LANL`s weapons research, development, and testing (WRD&T) program; special materials recovery; stockpile survelliance/evaluation; advanced fuels and heat sources development and production; and safe, secure storage of existing nuclear materials inventories. The problem is further exacerbated by LANL`s inability to ship any materials offsite because of the lack of receiver sites for mate rial and regulatory issues. Correction of the current deficiencies and enhancement of the facility will provide centralized storage close to a nuclear materials processing facility. The project will enable long-term, cost-effective storage in a secure environment with reduced radiation exposure to workers, and eliminate potential exposures to the public. This document provides Part I - Design Concept which describes the selected solution, and Part II - Project Management which describes the management system organization, the elements that make up the system, and the control and reporting system.

  7. Transformer design principles with applications to core-form power transformers

    CERN Document Server

    Del Vecchio, Robert M

    2010-01-01

    Updating and reorganizing the valuable information in the first edition to enhance logical development, Transformer Design Principles: With Applications to Core-Form Power Transformers, Second Edition remains focused on the basic physical concepts behind transformer design and operation. Starting with first principles, this book develops the reader's understanding of the rationale behind design practices by illustrating how basic formulae and modeling procedures are derived and used. Simplifies presentation and emphasizes fundamentals, making it easy to apply presented results to your own desi

  8. Design and pilot evaluation of the RAH-66 Comanche Core AFCS

    Science.gov (United States)

    Fogler, Donald L., Jr.; Keller, James F.

    1993-01-01

    This paper addresses the design and pilot evaluation of the Core Automatic Flight Control System (AFCS) for the Reconnaissance/Attack Helicopter (RAH-66) Comanche. During the period from November 1991 through February 1992, the RAH-66 Comanche control laws were evaluated through a structured pilot acceptance test using a motion base simulator. Design requirements, descriptions of the control law design, and handling qualities data collected from ADS-33 maneuvers are presented.

  9. Design of air-gapped magnetic-core inductors for superimposed direct and alternating currents

    Science.gov (United States)

    Ohri, A. K.; Wilson, T. G.; Owen, H. A., Jr.

    1976-01-01

    Using data on standard magnetic-material properties and standard core sizes for air-gap-type cores, an algorithm designed for a computer solution is developed which optimally determines the air-gap length and locates the quiescent point on the normal magnetization curve so as to yield an inductor design with the minimum number of turns for a given ac voltage and frequency and with a given dc bias current superimposed in the same winding. Magnetic-material data used in the design are the normal magnetization curve and a family of incremental permeability curves. A second procedure, which requires a simpler set of calculations, starts from an assigned quiescent point on the normal magnetization curve and first screens candidate core sizes for suitability, then determines the required turns and air-gap length.

  10. 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.

  11. Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design

    Science.gov (United States)

    Min, James B.; Ghosn, Louis J.; Lerch, Bradley A.; Raj, Sai V.; Holland, Frederic A., Jr.; Hebsur, Mohan G.

    2004-01-01

    The quest for cheap, low density and high performance materials in the design of aircraft and rotorcraft engine fan and propeller blades poses immense challenges to the materials and structural design engineers. The present study investigates the use of a sandwich foam fan blade mae up of solid face sheets and a metal foam core. The face sheets and the metal foam core material were an aerospace grade precipitation hardened 17-4 PH stainless steel with high strength and high toughness. The resulting structures possesses a high stiffness while being lighter than a similar solid construction. The material properties of 17-4 PH metal foam are reviewed briefly to describe the characteristics of sandwich structure for a fan blade application. A vibration analysis for natural frequencies and a detailed stress analysis on the 17-4 PH sandwich foam blade design for different combinations of kin thickness and core volume are presented with a comparison to a solid titanium blade.

  12. Core Noise: Implications of Emerging N+3 Designs and Acoustic Technology Needs

    Science.gov (United States)

    Hultgren, Lennart S.

    2011-01-01

    This presentation is a summary of the core-noise implications of NASA's primary N+3 aircraft concepts. These concepts are the MIT/P&W D8.5 Double Bubble design, the Boeing/GE SUGAR Volt hybrid gas-turbine/electric engine concept, the NASA N3-X Turboelectric Distributed Propulsion aircraft, and the NASA TBW-XN Truss-Braced Wing concept. The first two are future concepts for the Boeing 737/Airbus A320 US transcontinental mission of 180 passengers and a maximum range of 3000 nm. The last two are future concepts for the Boeing 777 transpacific mission of 350 passengers and a 7500 nm range. Sections of the presentation cover: turbofan design trends on the N+1.5 time frame and the already emerging importance of core noise; the NASA N+3 concepts and associated core-noise challenges; the historical trends for the engine bypass ratio (BPR), overall pressure ratio (OPR), and combustor exit temperature; and brief discussion of a noise research roadmap being developed to address the core-noise challenges identified for the N+3 concepts. The N+3 conceptual aircraft have (i) ultra-high bypass ratios, in the rage of 18 - 30, accomplished by either having a small-size, high-power-density core, an hybrid design which allows for an increased fan size, or by utilizing a turboelectric distributed propulsion design; and (ii) very high OPR in the 50 - 70 range. These trends will elevate the overall importance of turbomachinery core noise. The N+3 conceptual designs specify the need for the development and application of advanced liners and passive and active control strategies to reduce the core noise. Current engineering prediction of core noise uses semi-empirical methods based on older turbofan engines, with (at best) updates for more recent designs. The models have not seen the same level of development and maturity as those for fan and jet noise and are grossly inadequate for the designs considered for the N+3 time frame. An aggressive program for the development of updated noise

  13. Design of nuclear cells with re linking of trajectories; Diseno de celdas nucleares con re-encadenamiento de trayectorias

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, A.; Campos S, Y.; Ortiz S, J.J.; Montes, J.L.; Perrusquia, R.; Hernandez, J.L.; Torres, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2006-07-01

    ently work the results obtained with the Ohtli-RT system obtained when implementing the combinatory optimization technique well-known as Trajectories re linking or Path Re linking in English. The problem to solve is the radial design of nuclear fuel, taking like base nuclear fuel assembles for boiling water reactors (BWR Boiling Water Reactor by its initials in English). To evaluate the objective function used in the system the code in two dimensions Heliums 1.5 was used, which calculates the cross sections of the proposed design. The parameters that were considered for the evaluation of the objective function are the Power peak factor of the bar that generates bigger power in the cell and the Infinite Multiplication Factor. To prove the system its were used assembles 10x10 with 2 water channels. The obtained radial designs of nuclear fuel fulfilled the restrictions imposed to the considered limits, with regard to the involved parameters. (Author)

  14. Experimental determination of nuclear parameters for RP-0 reactor core; Determinacion experimental de los parametros nucleares para el nucleo tipo MTR del reactor nuclear RP-0

    Energy Technology Data Exchange (ETDEWEB)

    Cajacuri, Rafael A. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    2000-07-01

    In the nuclear reactor for investigations RP-0 which is in Lima, Peru, that is a open pool class reactor with 1 to 10 watts of power and as a nuclear fuel uranium 238 enriched to 20% constituted by elements of Material Testing Reactor fuel class. This has reflectors of graphite and moderator of water demineralized. In 1996/1997 was measured in this reactor the following parameters: position of the control bar that make critic the reactor, critic height of moderator, excess of reactivity of the nucleus, parameter of reactivity for vacuum, parameter of reactivity for temperature, reactivity of its control bar, levels of doses in the reactor. (author)

  15. Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

    Energy Technology Data Exchange (ETDEWEB)

    LECHELT, J.A.

    2000-10-17

    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.

  16. An introduction to the design, commissioning and operation of nuclear air cleaning systems for Qinshan Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Xinliang Chen; Jiangang Qu; Minqi Shi [Shanghai Nuclear Engineering Research and Design Institute (China)] [and others

    1995-02-01

    This paper introduces the design evolution, system schemes and design and construction of main nuclear air cleaning components such as HEPA filter, charcoal adsorber and concrete housing etc. for Qinshan 300MW PWR Nuclear Power Plant (QNPP), the first indigenously designed and constructed nuclear power plant in China. The field test results and in-service test results, since the air cleaning systems were put into operation 18 months ago, are presented and evaluated. These results demonstrate that the design and construction of the air cleaning systems and equipment manufacturing for QNPP are successful and the American codes and standards invoked in design, construction and testing of nuclear air cleaning systems for QNPP are applicable in China. The paper explains that the leakage rate of concrete air cleaning housings can also be assured if sealing measures are taken properly and embedded parts are designed carefully in the penetration areas of the housing and that the uniformity of the airflow distribution upstream the HEPA filters can be achieved generally no matter how inlet and outlet ducts of air cleaning unit are arranged.

  17. Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Ritterbusch, S.E.

    2000-08-01

    The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants.

  18. System and Software Design for the Plant Protection System for Shin-Hanul Nuclear Power Plant Units 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, In Seok; Kim, Young Geul; Choi, Woong Seock; Sohn, Se Do [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    The Reactor Protection System(RPS) protects the core fuel design limits and reactor coolant system pressure boundary for Anticipated Operational Occurrences (AOOs), and provides assistance in mitigating the consequences of Postulated Accidents (PAs). The ESFAS sends the initiation signals to Engineered Safety Feature - Component Control System (ESF-CCS) to mitigate consequences of design basis events. The Common Q platform Programmable Logic Controller (PLC) was used for Shin-Wolsung Nuclear Power Plant Units 1 and 2 and Shin-Kori Nuclear Power Plant Units 1, 2, 3 and 4 since Digital Plant Protection System (DPPS) based on Common Q PLC was applied for Ulchin Nuclear Power Plant Units 5 and 6. The PPS for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2) was developed using POSAFE-Q PLC for the first time for the PPS. The SHN1 and 2 PPS was delivered to the sites after completion of Man Machine Interface System Integrated System Test (MMIS-IST). The SHN1 and 2 PPS was developed to have the redundancy in each channel and to use the benefits of POSAFE-Q PLC, such as diagnostic and data communication. The PPS application software was developed using ISODE to minimize development time and human errors, and to improve software quality, productivity, and reusability.

  19. The design and performance of IceCube DeepCore

    Science.gov (United States)

    Abbasi, R.; Abdou, Y.; Abu-Zayyad, T.; Ackermann, M.; Adams, J.; Aguilar, J. A.; Ahlers, M.; Allen, M. M.; Altmann, D.; Andeen, K.; Auffenberg, J.; Bai, X.; Baker, M.; Barwick, S. W.; Bay, R.; Bazo Alba, J. L.; Beattie, K.; Beatty, J. J.; Bechet, S.; Becker, J. K.; Becker, K.-H.; Benabderrahmane, M. L.; BenZvi, S.; Berdermann, J.; Berghaus, P.; Berley, D.; Bernardini, E.; Bertrand, D.; Besson, D. Z.; Bindig, D.; Bissok, M.; Blaufuss, E.; Blumenthal, J.; Boersma, D. J.; Bohm, C.; Bose, D.; Böser, S.; Botner, O.; Brown, A. M.; Buitink, S.; Caballero-Mora, K. S.; Carson, M.; Chirkin, D.; Christy, B.; Clevermann, F.; Cohen, S.; Colnard, C.; Cowen, D. F.; Cruz Silva, A. H.; D'Agostino, M. V.; Danninger, M.; Daughhetee, J.; Davis, J. C.; De Clercq, C.; Degner, T.; Demirörs, L.; Descamps, F.; Desiati, P.; de Vries-Uiterweerd, G.; DeYoung, T.; Díaz-Vélez, J. C.; Dierckxsens, M.; Dreyer, J.; Dumm, J. P.; Dunkman, M.; Eisch, J.; Ellsworth, R. W.; Engdegård, O.; Euler, S.; Evenson, P. A.; Fadiran, O.; Fazely, A. R.; Fedynitch, A.; Feintzeig, J.; Feusels, T.; Filimonov, K.; Finley, C.; Fischer-Wasels, T.; Fox, B. D.; Franckowiak, A.; Franke, R.; Gaisser, T. K.; Gallagher, J.; Gerhardt, L.; Gladstone, L.; Glüsenkamp, T.; Goldschmidt, A.; Goodman, J. A.; Góra, D.; Grant, D.; Griesel, T.; Groß, A.; Grullon, S.; Gurtner, M.; Ha, C.; Haj Ismail, A.; Hallgren, A.; Halzen, F.; Han, K.; Hanson, K.; Heinen, D.; Helbing, K.; Hellauer, R.; Hickford, S.; Hill, G. C.; Hoffman, K. D.; Hoffmann, B.; Homeier, A.; Hoshina, K.; Huelsnitz, W.; Hülß, J.-P.; Hulth, P. O.; Hultqvist, K.; Hussain, S.; Ishihara, A.; Jacobi, E.; Jacobsen, J.; Japaridze, G. S.; Johansson, H.; Kampert, K.-H.; Kappes, A.; Karg, T.; Karle, A.; Kenny, P.; Kiryluk, J.; Kislat, F.; Klein, S. R.; Köhne, J.-H.; Kohnen, G.; Kolanoski, H.; Köpke, L.; Koskinen, D. J.; Kowalski, M.; Kowarik, T.; Krasberg, M.; Kroll, G.; Kurahashi, N.; Kuwabara, T.; Labare, M.; Laihem, K.; Landsman, H.; Larson, M. J.; Lauer, R.; Lünemann, J.; Madsen, J.; Marotta, A.; Maruyama, R.; Mase, K.; Matis, H. S.; Meagher, K.; Merck, M.; Mészáros, P.; Meures, T.; Miarecki, S.; Middell, E.; Milke, N.; Miller, J.; Montaruli, T.; Morse, R.; Movit, S. M.; Nahnhauer, R.; Nam, J. W.; Naumann, U.; Nygren, D. R.; Odrowski, S.; Olivas, A.; Olivo, M.; O'Murchadha, A.; Panknin, S.; Paul, L.; Pérez de los Heros, C.; Petrovic, J.; Piegsa, A.; Pieloth, D.; Porrata, R.; Posselt, J.; Price, P. B.; Przybylski, G. T.; Rawlins, K.; Redl, P.; Resconi, E.; Rhode, W.; Ribordy, M.; Richman, M.; Rodrigues, J. P.; Rothmaier, F.; Rott, C.; Ruhe, T.; Rutledge, D.; Ruzybayev, B.; Ryckbosch, D.; Sander, H.-G.; Santander, M.; Sarkar, S.; Schatto, K.; Schmidt, T.; Schönwald, A.; Schukraft, A.; Schultes, A.; Schulz, O.; Schunck, M.; Seckel, D.; Semburg, B.; Seo, S. H.; Sestayo, Y.; Seunarine, S.; Silvestri, A.; Spiczak, G. M.; Spiering, C.; Stamatikos, M.; Stanev, T.; Stezelberger, T.; Stokstad, R. G.; Stößl, A.; Strahler, E. A.; Ström, R.; Stüer, M.; Sullivan, G. W.; Swillens, Q.; Taavola, H.; Taboada, I.; Tamburro, A.; Tepe, A.; Ter-Antonyan, S.; Tilav, S.; Toale, P. A.; Toscano, S.; Tosi, D.; van Eijndhoven, N.; Vandenbroucke, J.; Van Overloop, A.; van Santen, J.; Vehring, M.; Voge, M.; Walck, C.; Waldenmaier, T.; Wallraff, M.; Walter, M.; Weaver, Ch.; Wendt, C.; Westerhoff, S.; Whitehorn, N.; Wiebe, K.; Wiebusch, C. H.; Williams, D. R.; Wischnewski, R.; Wissing, H.; Wolf, M.; Wood, T. R.; Woschnagg, K.; Xu, C.; Xu, D. L.; Xu, X. W.; Yanez, J. P.; Yodh, G.; Yoshida, S.; Zarzhitsky, P.; Zoll, M.

    2012-05-01

    The IceCube neutrino observatory in operation at the South Pole, Antarctica, comprises three distinct components: a large buried array for ultrahigh energy neutrino detection, a surface air shower array, and a new buried component called DeepCore. DeepCore was designed to lower the IceCube neutrino energy threshold by over an order of magnitude, to energies as low as about 10 GeV. DeepCore is situated primarily 2100 m below the surface of the icecap at the South Pole, at the bottom center of the existing IceCube array, and began taking physics data in May 2010. Its location takes advantage of the exceptionally clear ice at those depths and allows it to use the surrounding IceCube detector as a highly efficient active veto against the principal background of downward-going muons produced in cosmic-ray air showers. DeepCore has a module density roughly five times higher than that of the standard IceCube array, and uses photomultiplier tubes with a new photocathode featuring a quantum efficiency about 35% higher than standard IceCube PMTs. Taken together, these features of DeepCore will increase IceCube's sensitivity to neutrinos from WIMP dark matter annihilations, atmospheric neutrino oscillations, galactic supernova neutrinos, and point sources of neutrinos in the northern and southern skies. In this paper we describe the design and initial performance of DeepCore.

  20. Understanding the selection of core head design features to match precisely challenging well applications

    Energy Technology Data Exchange (ETDEWEB)

    Zambrana, Roberto; Sousa, J. Tadeu V. de; Antunes, Ricardo [Halliburton Servicos Ltda., Rio de Janeiro, RJ (Brazil)

    2008-07-01

    Reliable rock mechanical information is very important for optimum reservoir development. This information can help specialists to accurately estimate reserves, reservoir compaction, sand production, stress field orientation, etc. In all cases, the solutions to problems involving rock mechanics lead to significant cost savings. Consequently, it is important that the decisions be based on the most accurate information possible. For the describing rock mechanics, cores represent the major source of data and therefore should be of good quality. However, there are several well conditions that cause coring and core recovery to be difficult, for example: unconsolidated formations; laminated and fractured rocks; critical mud losses, etc. The problem becomes even worse in high-inclination wells with long horizontal sections. In such situations, the optimum selections of core heads become critical. This paper will discuss the most important design features that enable core heads to be matched precisely to various challenging applications. Cases histories will be used to illustrate the superior performance of selected core heads. They include coring in horizontal wells and in harsh well conditions with critical mud losses. (author)

  1. Geoantineutrino Spectrum and Slow Nuclear Burning on the Boundary of the Liquid and Solid Phases of the Earth's core

    CERN Document Server

    Rusov, V D; Khotyaintseva, E N; Kosenko, S I; Litvinov, D A; Pavlovich, V N; Tarasov, V A; Vaschenko, V N; Zelentsova, T N

    2004-01-01

    The problem of the geoantineutrino deficit and the experimental results of the interaction of uranium dioxide and carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600- 22000 C) have induced us to consider the possible consequences of made by V. Anisichkin and A. Ershov supposition that there is an actinoid shell on boundary of liquid and solid phases of the Earth's core. We have shown that the activation of a natural nuclear reactor operating as the solitary waves of nuclear burning in 238U- and/or 232Th-medium (in particular, the neutron-fission progressive wave of Feoktistov and/or Teller-Ishikawa-Wood) such physical consequent can be. The simplified model of the kinetics of accumulation and burnup in U-Pu fuel cycle of Feoktistov is developed. The results of the numerical simulation of neutron-fission wave in two-phase UO2/Fe medium on a surface of the Earth's solid core are presented. On the basis of O'Nions-Ivensen-Hamilton model of the geochemical evolution...

  2. Application of the nuclear equation of state obtained by the variational method to core-collapse supernovae

    CERN Document Server

    Togashi, H; Sumiyoshi, K; Nakazato, K

    2014-01-01

    The equation of state (EOS) for hot asymmetric nuclear matter which is constructed with the variational method starting from the Argonne v18 and Urbana IX nuclear forces is applied to spherically symmetric core-collapse supernovae (SNe). We first investigate the EOS of isentropic beta-stable SN matter, and find that the matter with the variational EOS is more neutron-rich than that with the Shen EOS. Using the variational EOS for uniform matter supplemented by the Shen EOS of non-uniform matter at low densities, we perform general-relativistic spherically symmetric simulations of core-collapse SNe with and without neutrino transfer, starting from a presupernova model of 15 solar mass. In the adiabatic simulation without neutrino transfer, the explosion is successful, and the explosion energy with the variational EOS is larger than that with the Shen EOS. In the case of the simulation with neutrino transfer, the shock wave stalls and then the explosion fails, as in other spherically symmetric simulations. The ...

  3. Minimization of the energy loss of nuclear power plants in case of partial in-core monitoring system failure

    Science.gov (United States)

    Zagrebaev, A. M.; Ramazanov, R. N.; Lunegova, E. A.

    2017-01-01

    In this paper we consider the optimization problem minimize of the energy loss of nuclear power plants in case of partial in-core monitoring system failure. It is possible to continuation of reactor operation at reduced power or total replacement of the channel neutron measurements, requiring shutdown of the reactor and the stock of detectors. This article examines the reconstruction of the energy release in the core of a nuclear reactor on the basis of the indications of height sensors. The missing measurement information can be reconstructed by mathematical methods, and replacement of the failed sensors can be avoided. It is suggested that a set of ‘natural’ functions determined by means of statistical estimates obtained from archival data be constructed. The procedure proposed makes it possible to reconstruct the field even with a significant loss of measurement information. Improving the accuracy of the restoration of the neutron flux density in partial loss of measurement information to minimize the stock of necessary components and the associated losses.

  4. Design of the Face/Core Interface for Improved Fracture Resistance

    DEFF Research Database (Denmark)

    Lundsgaard-Larsen, Christian; Berggreen, Christian; Carlsson, Leif A.

    2008-01-01

    This study investigates the face/core fracture behavior of sandwich specimens with different designs. The traditional interface with a quadraxial mat directly adhered to the foam core is compared to interfaces where an additional mat with randomly oriented fibers is inserted between core and face....... The extra mat affects the crack propagation path in the sandwich specimen, and makes it more likely for the crack to propagate at or near the interface, instead of kinking into the laminate or core. Further, the extra mat acts as a source for fiber bridging, and hereby the fracture resistance is increased...... as bridging fibers shield the crack tip from the loading. Results show that the increase in fracture resistance due to fiber bridging is significant. Cohesive laws regarding cracking of sandwich interfaces are extracted....

  5. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in a...

  6. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in...

  7. Optimal hardware/software co-synthesis for core-based SoC designs

    Institute of Scientific and Technical Information of China (English)

    Zhan Jinyu; Xiong Guangze

    2006-01-01

    A hardware/software co-synthesis method is presented for SoC designs consisting of both hardware IP cores and software components on a graph-theoretic formulation. Given a SoC integrated with a set of functions and a set of performance factors, a core for each function is selected from a set of alternative IP cores and software components, and optimal partitions is found in a way to evenly balance the performance factors and to ultimately reduce the overall cost, size, power consumption and runtime of the core-based SoC. The algorithm formulates IP cores and components into the corresponding mathematical models, presents a graph-theoretic model for finding the optimal partitions of SoC design and transforms SoC hardware/software co-synthesis problem into finding optimal paths in a weighted, directed graph. Overcoming the three main deficiencies of the traditional methods, this method can work automatically, evaluate more performance factors at the same time and meet the particularity of SoC designs.At last, the approach is illustrated that is practical and effective through partitioning a practical system.

  8. Nuclear Electric Vehicle Optimization Toolset (NEVOT): Integrated System Design Using Genetic Algorithms

    Science.gov (United States)

    Tinker, Michael L.; Steincamp, James W.; Stewart, Eric T.; Patton, Bruce W.; Pannell, William P.; Newby, Ronald L.; Coffman, Mark E.; Qualls, A. L.; Bancroft, S.; Molvik, Greg

    2003-01-01

    The Nuclear Electric Vehicle Optimization Toolset (NEVOT) optimizes the design of all major Nuclear Electric Propulsion (NEP) vehicle subsystems for a defined mission within constraints and optimization parameters chosen by a user. The tool uses a Genetic Algorithm (GA) search technique to combine subsystem designs and evaluate the fitness of the integrated design to fulfill a mission. The fitness of an individual is used within the GA to determine its probability of survival through successive generations in which the designs with low fitness are eliminated and replaced with combinations or mutations of designs with higher fitness. The program can find optimal solutions for different sets of fitness metrics without modification and can create and evaluate vehicle designs that might never be conceived of through traditional design techniques. It is anticipated that the flexible optimization methodology will expand present knowledge of the design trade-offs inherent in designing nuclear powered space vehicles and lead to improved NEP designs.

  9. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  10. Comparative Evaluation of a CORE Based Learning Environment for Nuclear Medicine.

    Science.gov (United States)

    Hogg, Peter; Boyle, Tom; Lawson, Richard

    1999-01-01

    Reports on a comparative assessment of a multimedia learning environment based on a guided discovery approach called CORE (Concept Object Refinement Expression) with two control conditions, lecture and electronic book, in an undergraduate radiography course. Discusses results of qualitative and quantitative measures of effectiveness, pretests and…

  11. Generic approach for designing and implementing a passive autocatalytic recombiner PAR-system in nuclear power plant containments

    Energy Technology Data Exchange (ETDEWEB)

    Bachellerie, E. E-mail: bachelle@tecatom.fr; Arnould, F.; Auglaire, M.; Boeck, B. de; Braillard, O.; Eckardt, B.; Ferroni, F.; Moffett, R

    2003-04-01

    PARSOAR project is one of the scientific projects on hydrogen risk in nuclear power plants co-sponsored by the European Commission in the Fifth Euratom Framework Program and the Swiss government. It is one of the components of the 'Safety Accident Management' (SAM) cluster that is focusing on review of severe accident risks and on development of countermeasures for defence-in-depth. The first objective is to carry out a state-of-the-art on passive autocatalytic recombiners (PAR) to constitute a database helpful for the PAR-designers, nuclear power plant designers and utilities, and safety authorities. The second objective is to elaborate a handbook guide that defines an approach for implementing catalytic recombiners in nuclear power plants, and that proposes recommended practices to support the detailed implementation by individual nuclear actors. The third objective is to identify the needs in terms of complementary experimental qualification or computer code developments about the hydrogen risk in nuclear power plants. It is also recommended that future experimental tests and numerical modelling should concentrate on the remaining issues of (a) lowering of existing uncertainties about hydrogen sources as core reflood or the reactions between boron carbide and steam or uranium and steam, (b) completing the experimental qualification of catalytic recombiners for existing and future nuclear power plant applications, and (c) performing well-qualified couplings between computational fluid dynamics computer codes and numerical models simulating PAR-behaviour (thanks to some global and well-instrumented experiments reproducing a complete reference accident sequence in presence of catalytic recombiners). Such areas to be considered for future work could be integrated in the scope of a European hydrogen network of excellence, in the framework of the Sixth Euratom Framework Program (2002-2006)

  12. Lack of nuclear clusters in dwarf spheroidal galaxies: implications for massive black holes formation and the cusp/core problem

    Science.gov (United States)

    Arca-Sedda, Manuel; Capuzzo-Dolcetta, Roberto

    2017-01-01

    One of the leading scenarios for the formation of nuclear star clusters in galaxies is related to the orbital decay of globular clusters (GCs) and their subsequent merging, though alternative theories are currently debated. The availability of high-quality data for structural and orbital parameters of GCs allows us to test different nuclear star cluster formation scenarios. The Fornax dwarf spheroidal (dSph) galaxy is the heaviest satellite of the Milky Way and it is the only known dSph hosting five GCs, whereas there are no clear signatures for the presence of a central massive black hole. For this reason, it represents a suited place to study the orbital decay process in dwarf galaxies. In this paper, we model the future evolution of the Fornax GCs simulating them and the host galaxy by means of direct N-body simulations. Our simulations also take into account the gravitational field generated by the Milky Way. We found that if the Fornax galaxy is embedded in a standard cold dark matter halo, the nuclear cluster formation would be significantly hampered by the high central galactic mass density. In this context, we discuss the possibility that infalling GCs drive the flattening of the galactic density profile, giving a possible alternative explanation to the so-called cusp/core problem. Moreover, we briefly discuss the link between GC infall process and the absence of massive black holes in the centre of dSphs.

  13. Thermodynamic evaluation of the solidification phase of molten core-concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

    Science.gov (United States)

    Kitagaki, Toru; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro

    2017-04-01

    The solidification phases of molten core-concrete under the estimated molten core-concrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool, FactSage 6.2, and the NUCLEA database in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core-concrete forms (U,Zr)O2 and (Zr,U)SiO4, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO4 requires a relatively long time because it involves a change in the crystal structure from fcc-(U,Zr)O2 to tet-(U,Zr)O2, followed by the formation of (Zr,U)SiO4 by reaction with SiO2. Therefore, the formation of (Zr,U)SiO4 is limited under quenching conditions. Other common phases are the oxide phases, CaAl2Si2O8, SiO2, and CaSiO3, and the metallic phases of the Fe-Si and Fe-Ni alloys. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

  14. Special Sm core complex functions in assembly of the U2 small nuclear ribonucleoprotein of Trypanosoma brucei.

    Science.gov (United States)

    Preusser, Christian; Palfi, Zsofia; Bindereif, Albrecht

    2009-08-01

    The processing of polycistronic pre-mRNAs in trypanosomes requires the spliceosomal small ribonucleoprotein complexes (snRNPs) U1, U2, U4/U6, U5, and SL, each of which contains a core of seven Sm proteins. Recently we reported the first evidence for a core variation in spliceosomal snRNPs; specifically, in the trypanosome U2 snRNP, two of the canonical Sm proteins, SmB and SmD3, are replaced by two U2-specific Sm proteins, Sm15K and Sm16.5K. Here we identify the U2-specific, nuclear-localized U2B'' protein from Trypanosoma brucei. U2B'' interacts with a second U2 snRNP protein, U2-40K (U2A'), which in turn contacts the U2-specific Sm16.5K/15K subcomplex. Together they form a high-affinity, U2-specific binding complex. This trypanosome-specific assembly differs from the mammalian system and provides a functional role for the Sm core variation found in the trypanosomal U2 snRNP.

  15. Designing a nuclear data base prototype using Oracle and Prolog

    Energy Technology Data Exchange (ETDEWEB)

    Paviotti-Corcuera, R.; Ford, C.E.; Perez, R.B.

    1988-11-01

    An ever-increasing demand exists for easily accessible nuclear data base systems. The purpose of this work is to analyze the feasibility of using artificial intelligence methods as tools to provide the necessary functionality to extract information from nuclear data files in a user-friendly manner. For the prototype of this work, a sample of data that can be later enlarged to a complete, evaluated nuclear data base has been used. To implement this prototype, two approaches have been followed: a conventional approach using the commercially available Oracle relational data base management system; and an artificial intelligence approach using the Prolog programming language. This prototypic work shows the feasibility of applying artificial intelligence methods to data bases, and represents a first step toward development of intelligent nuclear data base systems. The characteristics of the query language from both approaches make the second one preferable from a user's point of view. 23 refs., 7 tabs.

  16. Improving human reliability through better nuclear power plant system design: Program for advanced nuclear power studies. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Golay, M.W.

    1993-10-10

    The project on ``Development of a Theory of the Dependence of Human Reliability upon System Designs as a Means of Improving Nuclear Power Plant Performance`` was been undertaken in order to address the problem of human error in advanced nuclear power plant designs. Lack of a mature theory has retarded progress in reducing likely frequencies of human errors. Work being pursued in this project is to perform a set of experiments involving human subjects who are required to operate, diagnose and respond to changes in computer-simulated systems, relevant to those encountered in nuclear power plants, which are made to differ in complexity in a systematic manner. The computer program used to present the problems to be solved also records the response of the operator as it unfolds.

  17. Integrating IMS Learning Design and IMS Question and Test Interoperability using CopperCore Service Integration

    NARCIS (Netherlands)

    Vogten, Hubert; Martens, Harrie; Nadolski, Rob; Tattersall, Colin; Van Rosmalen, Peter; Koper, Rob

    2006-01-01

    Please, cite this publication as: Vogten, H., Martens, H., Nadolski, R., Tattersall, C., van Rosmalen, P., & Koper, R. (2006). Integrating IMS Learning Design and IMS Question and Test Interoperability using CopperCore Service Integration. Proceedings of International Workshop in Learning Networks f

  18. CopperCore Service Integration, Integrating IMS Learning Design and IMS Question and Test Interoperability

    NARCIS (Netherlands)

    Vogten, Hubert; Martens, Harrie; Nadolski, Rob; Tattersall, Colin; Van Rosmalen, Peter; Koper, Rob

    2006-01-01

    Vogten, H., Martens, H., Nadolski, R., Tattersall, C., Rosmalen, van, P., Koper, R., (2006). CopperCore Service Integration, Integrating IMS Learning Design and IMS Question and Test Interoperability. Proceedings of the 6th IEEE International Conference on Advanced Learning Technologies (pp. 378-379

  19. CopperCore: a service based approach towards implementing the IMS Learning Design specification.

    NARCIS (Netherlands)

    Vogten, Hubert

    2006-01-01

    This paper presents a service developed by the Open University of the Netherlands, called CopperCore which implements an IMS Learning Design engine as service. The overall architecture is described including a detailed description of the web service application programming interfaces.

  20. CopperCore Service Integration, Integrating IMS Learning Design and IMS Question and Test Interoperability

    NARCIS (Netherlands)

    Vogten, Hubert; Martens, Harrie; Nadolski, Rob; Tattersall, Colin; Van Rosmalen, Peter; Koper, Rob

    2006-01-01

    Vogten, H., Martens, H., Nadolski, R., Tattersall, C., Rosmalen, van, P., Koper, R., (2006). CopperCore Service Integration, Integrating IMS Learning Design and IMS Question and Test Interoperability. Proceedings of the 6th IEEE International Conference on Advanced Learning Technologies (pp.

  1. CopperCore: a service based approach towards implementing the IMS Learning Design specification.

    NARCIS (Netherlands)

    Vogten, Hubert

    2006-01-01

    This paper presents a service developed by the Open University of the Netherlands, called CopperCore which implements an IMS Learning Design engine as service. The overall architecture is described including a detailed description of the web service application programming interfaces.

  2. Small core Chalcogenide photonic crystal fiber for midinfrared wavelength conversion: experiment and design

    OpenAIRE

    Xing, Sida; Grassani, Davide; Kharitonov, Svyatoslav; Billat, Adrien; Brès, Camille-Sophie

    2016-01-01

    Kerr index and dispersion parameter of a small core chalcogenide photonic crystal fiber are estimated via four-wave mixing near 2μm. From these values, new fiber design is proposed to efficiently generate idlers in mid-infrared.

  3. CopperCore Service Integration, Integrating IMS Learning Design and IMS Question and Test Interoperability

    NARCIS (Netherlands)

    Vogten, Hubert; Martens, Harrie; Nadolski, Rob; Tattersall, Colin; Van Rosmalen, Peter; Koper, Rob

    2006-01-01

    Vogten, H., Martens, H., Nadolski, R., Tattersall, C., Rosmalen, van, P., Koper, R., (2006). CopperCore Service Integration, Integrating IMS Learning Design and IMS Question and Test Interoperability. Proceedings of the 6th IEEE International Conference on Advanced Learning Technologies (pp. 378-379

  4. Light source design using Kagome-lattice hollow core photonic crystal fibers

    Science.gov (United States)

    Hossain, Md. Anwar; Namihira, Yoshinori

    2014-09-01

    Supercontinuum (SC) light source is designed using high pressure Xe-filled hollow core Kagome-lattice photonic crystal fiber. Using finite element method with perfectly matched layer, SC spectra in normal chromatic dispersion region have been generated using picosecond optical pulses from relatively less expensive laser sources.

  5. Design Guideline of Hollow-Core Fibres with Cobweb Cladding Structure

    Institute of Scientific and Technical Information of China (English)

    HUO Liang; YU Rong-Jin; ZHANG Bing; CHEN Ming-Yang; LI Bing-Xin

    2006-01-01

    @@ By using a plane wave expansion method, some important parameters of designing the hollow-core fibre with cobweb cladding structure are analysed. Taking a dielectric material PMMA, for example, the tolerance of the parameters is discussed. The results show that the parameters of the structure possess oneselfofa regularity and limit, and have a larger tolerance for the structural parameters in fabrication.

  6. Designing high frequency ac inductors using ferrite and Molypermalloy Powder Cores (MPP)

    Science.gov (United States)

    Mclyman, W. T.; Wagner, A. P.

    1985-01-01

    The major considerations in the design of high frequency ac inductors are reviewed. Two methods for designing the inductor: the area product method and the core geometry method, are presented. The two major effects of the inductor air gap, fringing flux power loss and increase of inductance, are discussed. Equations for the inductor design and a step-by-step design procedure are given. The use of a lumped air gap or a distributed air gap are discussed and a comparison of the losses resulting from these gaps, together with experimental results are presented.

  7. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    Science.gov (United States)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  8. An overview of the FZJ-tools for HTR core design and reactor dynamics, the past, present and future

    Energy Technology Data Exchange (ETDEWEB)

    Reitsma, F. [Nuclear Engineering Analysis (NEA), PBMR (Pty Ltd), cennturion (South Africa); Rutten, H.J.; Scherer, W. [Forschungzentrum Julich GmbH, Institute for Safety Research and Reactor Technology, Julich (Germany)

    2005-07-01

    The development of the pebble-bed type high-temperature reactor in Germany was actively supported by the research centre Juelich (FZJ, former KFA) i.e. with the development of theoretical methods and computational tools to perform core neutronics design, reactor operation simulation and transient analysis. The tools, developed as the outcome of research activities, made huge contributions not only to the understanding of the technology and its physical behaviour but were also used in support of licensing of German HTR projects. Today these codes are used in the design and licensing of current commercial projects such as the Pebble Bed Modular Reactor (PBMR) but is also used to design the upcoming Gen IV reactors of HTR pebble-bed type. The renewed interest in these codes especially with respect to the pebble-bed designs is due to their unique features such as the fuel management algorithms, the simultaneous treatment of nuclear, thermal-hydraulic and fluid-dynamic problems and the description of fast and long-term transients. The paper provides an overview of the codes VSOP and TINTE, provides an update of recent developments in the codes and gives specific examples of applications such as the PBMR 400 MW running-in phase, a comparison with the SANA pebble bed effective thermal conductivity experiment and very recent results obtained simulating the corrosion due to an air ingress event simulated in the NACOK facility. Finally some ideas on the future development of these codes are discussed. (authors)

  9. Human factors design review guidelines for advanced nuclear control room technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, J.; Brown, W. (Brookhaven National Lab., Upton, NY (United States)); Granda, T.; Baker, C. (Carlow Associates, Inc., Fairfax, VA (United States))

    1991-01-01

    Advanced control rooms (ACRs) for future nuclear power plants are being designed utilizing computer-based technologies. The US Nuclear Regulatory Commission reviews the human engineering aspects of such control rooms to ensure that they are designed to good human factors engineering principles and that operator performance and reliability are appropriately supported in order to protect public health and safety. This paper describes the rationale, general approach, and initial development of an NRC Advanced Control Room Design Review Guideline. 20 refs., 1 fig.

  10. An algorithm for multi-group two-dimensional neutron diffusion kinetics in nuclear reactor cores

    OpenAIRE

    Marcelo Schramm

    2016-01-01

    The objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neutron diffusion kinetics in a reactor core. The presented methodology uses a polyno- mial approximation in a rectangular homogeneous domain with non{homogeneous boundary conditions. As it consists on a truncated Taylor series, its error estimates varies with the size of the rectangle. The coefficients are obtained mainly by their relations with the independent term, which is determined by the dif...

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

    CERN Document Server

    Konieczka, Maciej

    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.

  12. Neutron stars with hyperon cores: stellar radii and EOS near nuclear density

    CERN Document Server

    Fortin, M; Haensel, P; Bejger, M

    2014-01-01

    The existence of 2 Msun pulsars puts very strong constraints on the equation of state (EOS) of neutron stars (NSs) with hyperon cores, which can be satisfied only by special models of hadronic matter. The radius-mass relation for these models is so specific that it could be submitted to an observational test with forthcoming X-ray observatories. We want to study the impact of the presence of hyperon cores on the radius-mass relation for NS. We aim at finding how, and for which particular stellar mass range, a specific relation R(M), where M is gravitational mass, and R is radius, is associated with the presence of an hyperon core. We consider a large set of theoretical EOSs of dense matter, based on the relativistic mean-field (RMF) approximation, allowing for the presence of hyperons in NSs. We seek for correlations between R(M) and the stiffness of the EOS below the hyperon threshold, needed to pass the 2 Msun test. For NS masses 1.013km, which is due to a very stiff pre-hyperon segment of the EOS. At nucle...

  13. Nuclear Thermal Rocket/Vehicle Design Options for Future NASA Missions to the Moon and Mars

    Science.gov (United States)

    Borowski, Stanley K.; Corban, Robert R.; Mcguire, Melissa L.; Beke, Erik G.

    1995-01-01

    The nuclear thermal rocket (NTR) provides a unique propulsion capability to planners/designers of future human exploration missions to the Moon and Mars. In addition to its high specific impulse (approximately 850-1000 s) and engine thrust-to-weight ratio (approximately 3-10), the NTR can also be configured as a 'dual mode' system capable of generating electrical power for spacecraft environmental systems, communications, and enhanced stage operations (e.g., refrigeration for long-term liquid hydrogen storage). At present the Nuclear Propulsion Office (NPO) is examining a variety of mission applications for the NTR ranging from an expendable, single-burn, trans-lunar injection (TLI) stage for NASA's First Lunar Outpost (FLO) mission to all propulsive, multiburn, NTR-powered spacecraft supporting a 'split cargo-piloted sprint' Mars mission architecture. Each application results in a particular set of requirements in areas such as the number of engines and their respective thrust levels, restart capability, fuel operating temperature and lifetime, cryofluid storage, and stage size. Two solid core NTR concepts are examined -- one based on NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor (NDR) technology, and a second concept which utilizes a ternary carbide 'twisted ribbon' fuel form developed by the Commonwealth of Independent States (CIS). The NDR and CIS concepts have an established technology database involving significant nuclear testing at or near representative operating conditions. Integrated systems and mission studies indicate that clusters of two to four 15 to 25 klbf NDR or CIS engines are sufficient for most of the lunar and Mars mission scenarios currently under consideration. This paper provides descriptions and performance characteristics for the NDR and CIS concepts, summarizes NASA's First Lunar Outpost and Mars mission scenarios, and describes characteristics for representative cargo and piloted vehicles compatible with a

  14. Nuclear Thermal Rocket/vehicle design options for future NASA missions to the Moon and Mars

    Science.gov (United States)

    Borowski, Stanley K.; Corban, Robert R.; McGuire, Melissa L.; Beke, Erik G.

    1995-09-01

    The nuclear thermal rocket (NTR) provides a unique propulsion capability to planners/designers of future human exploration missions to the Moon and Mars. In addition to its high specific impulse (approximately 850-1000 s) and engine thrust-to-weight ratio (approximately 3-10), the NTR can also be configured as a 'dual mode' system capable of generating electrical power for spacecraft environmental systems, communications, and enhanced stage operations (e.g., refrigeration for long-term liquid hydrogen storage). At present the Nuclear Propulsion Office (NPO) is examining a variety of mission applications for the NTR ranging from an expendable, single-burn, trans-lunar injection (TLI) stage for NASA's First Lunar Outpost (FLO) mission to all propulsive, multiburn, NTR-powered spacecraft supporting a 'split cargo-piloted sprint' Mars mission architecture. Each application results in a particular set of requirements in areas such as the number of engines and their respective thrust levels, restart capability, fuel operating temperature and lifetime, cryofluid storage, and stage size. Two solid core NTR concepts are examined -- one based on NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor (NDR) technology, and a second concept which utilizes a ternary carbide 'twisted ribbon' fuel form developed by the Commonwealth of Independent States (CIS). The NDR and CIS concepts have an established technology database involving significant nuclear testing at or near representative operating conditions. Integrated systems and mission studies indicate that clusters of two to four 15 to 25 klbf NDR or CIS engines are sufficient for most of the lunar and Mars mission scenarios currently under consideration. This paper provides descriptions and performance characteristics for the NDR and CIS concepts, summarizes NASA's First Lunar Outpost and Mars mission scenarios, and describes characteristics for representative cargo and piloted vehicles compatible with a

  15. Study on Site Specific Design Earthquake Ground Motion of Nuclear Power Plants in China1

    Institute of Scientific and Technical Information of China (English)

    Zhou Bochang; Li Xiaojun; Li Yaqi

    2008-01-01

    The main technical backgrounds and requirements are introduced with regard to earthquake ground motion design parameters in several domestic and American standards,codes and guides involved in the seismic analysis and design activities of nuclear power plants in China.Based on the research results from site seismic safety evaluation of domestic nuclear power plant projects in the last years,characteristics and differences of site specific design spectra are analyzed in comparison with standard response spectra,and the suitability of standard response spectra for domestic nuclear power plant projects is discussed.

  16. Systematic approach for designing zero-DGD coupled multi-core optical fibers.

    Science.gov (United States)

    Parto, Midya; Eftekhar, Mohammad Amin; Miri, Mohammad-Ali; Amezcua-Correa, Rodrigo; Li, Guifang; Christodoulides, Demetrios N

    2016-05-01

    An analytical method is presented for designing N-coupled multi-core fibers with zero differential group delay. This approach effectively reduces the problem to a system of N-1 algebraic equations involving the associated coupling coefficients and propagation constants, as obtained from coupled mode theory. Once the parameters of one of the cores are specified, the roots of the resulting N-1 equations can be used to determine the characteristics of the remaining waveguide elements. Using this technique, a number of pertinent geometrical configurations are investigated to minimize intermodal dispersion.

  17. A systematic approach for designing zero-DGD coupled multi-core optical fibers

    CERN Document Server

    Parto, Midya; Miri, Mohammad-Ali; Amezcua-Correa, Rodrigo; Li, Guifang; Christodoulides, Demetrios N

    2016-01-01

    An analytical method is presented for designing N-coupled multi-core fibers with zero differential group delay. This approach effectively reduces the problem to a system of N-1 algebraic equations involving the associated coupling coefficients and propagation constants as obtained from coupled mode theory. Once the parameters of one of the cores are specified, the roots of the resulting N-1 equations can then be used to determine the characteristics of the remaining waveguide elements. Using this technique, a number of pertinent geometrical configurations are investigated in order to minimize intermodal dispersion.

  18. Use of nuclear data for space and aeronautic designs

    Energy Technology Data Exchange (ETDEWEB)

    Palau, M.C.; Carriere, Th. [Astrium ST, 78 - Les Mureaux (France); Buard, B.; Weulersse, C. [EADS IN, 92 - Suresnes (France); Saigne, F. [CEM2, 34 - Montpellier (France); Wrobel, F. [Nice Univ., LPES (France)

    2008-07-01

    Until recently, the effects of radiation environment on on-board electronics on launchers and aircraft had not been seriously taken into account. The situation has changed. And one of the most significant effects observed on on-board electronics is what we call Single Event Upset (SEU). This talk explains how the combination of electrical sensitivity of components and nuclear physics is important in the calculation of SEU rates, and emphasizes the aspects of nuclear physics useful to give the probability for a dangerous event to occur. Some circumvention methods will be rapidly identified. (authors)

  19. Fault-Tolerant Design and Testing of USB2.0 Peripheral Devices IP Core System

    Institute of Scientific and Technical Information of China (English)

    BAI Xiaoping; WEI Yuanfeng

    2007-01-01

    Universal serial bus 2.0 (USB2.0) is a kind of mainstream interface technology. The traditional USB developing is only to develop USB peripheral devices. For the USB2.0 peripheral devices IP core system that has wide application foreground, some interference inevitably exists in signal transmitting. Some fault-tolerant design and test methods must be adopted in order to correctly transmit and receive data. Combining with a project, this paper introduces in detail about measures, hardware implement, and test methods of fault-tolerant design about USB2.0 peripheral devices IP core system. Fault-tolerant design measures, noise reduction measures of signal processing, fault-tolerant methods about data encode and decode, package identification (ID) field fault-tolerant methods, and cyclic redundancy checks fault-tolerant methods are discussed. The paper also presents some hardware implement methods about fault-tolerant design of data decode and test methods about fault-tolerant design of USB2.0 IP core system. These methods can offer the reference for development of USB2.0 system in all kinds of electronics instrumentations.

  20. Statistical analysis in the design of nuclear fuel cells; Analisis estadistico en el diseno de celdas de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    This work presents the preliminary results of a statistical analysis carried out for the design of nuclear fuel cells. The analysis consists in verifying the behavior of a cell, related with the frequency of the pines used for its design. In this preliminary study was analyzed the behavior of infinite multiplication factor and the peak factor of local power. On the other hand, the mentioned analysis was carried out using a pines group of enriched uranium previously established, for which varies the pines frequency used in the design. To carry out the study, the CASMO-IV code was used. The obtained designs are for the different axial areas of a fuel assembly. A balance cycle of the unit 1 of the nuclear power plant of Laguna Verde was used like reference. To obtain the result of the present work, systems that are already had and in which have already been implemented the heuristic techniques of ant colonies, neural networks and a hybrid between the dispersed search and the trajectories re-chaining. The results show that is possible to design nuclear fuel cells with a good performance, if is considered a statistical behavior in the frequency of the used pines, in a same way. (Author)

  1. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  2. Reactor physics analysis for the design of nuclear fuel lattices with burnable poisons

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico); Guzman, Juan R., E-mail: maestro_juan_rafael@hotmail.com [Departamento de Fisica y Matematicas, Instituto Politecnico Nacional, Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Mexico, D.F. (Mexico)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer A fuel rod optimization for the coupled bundle-core design in a BWR is developed. Black-Right-Pointing-Pointer An algorithm to minimize the rod power peaking factor is used. Black-Right-Pointing-Pointer The fissile content is divided in two factors. Black-Right-Pointing-Pointer A reactor physics analysis of these factors is performed. Black-Right-Pointing-Pointer The algorithm is applied to a typical BWR fuel lattice. - Abstract: The main goals in nuclear fuel lattice design are: (1) minimizing the rod power peaking factor (PPF) in order that the power level distribution is the most uniform; (2) obtaining a prescribed target value for the multiplication factor (k) at the end of the irradiation in order that the fuel lattice reaches the desired reactivity; and (3) obtaining a prescribed target value for the k at the beginning of the irradiation in order that the reactivity excess is neither a high value (to ease the maneuvering of the control systems) nor a low value (to avoid the penalization of the high cost of the burnable poison content). In this work a simple algorithm to design the burnable poison bearing nuclear fuel lattice is presented. This algorithm is based on a reactor physics analysis. The algorithm is focused on finding the radial distribution of the fuel rods having different fissile and burnable poison contents in order to obtain: (1) an adequate minimum PPF; (2) a prescribed target value of the k at the end of the irradiation; and (3) a prescribed target value of the k at the beginning of the irradiation. This algorithm is based on the factorization of the fissile and burnable poison contents of each fuel rod and on the application of the first-order perturbation theory. The performance of the algorithm is demonstrated with the design of a fuel lattice composed of uranium dioxide (UO{sub 2}) and gadolinium dioxide (Gd{sub 2}O{sub 3}) for boiling water reactors (BWR). This algorithm has been accomplished

  3. Design Features and Technology Uncertainties for the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    John M. Ryskamp; Phil Hildebrandt; Osamu Baba; Ron Ballinger; Robert Brodsky; Hans-Wolfgang Chi; Dennis Crutchfield; Herb Estrada; Jeane-Claude Garnier; Gerald Gordon; Richard Hobbins; Dan Keuter; Marilyn Kray; Philippe Martin; Steve Melancon; Christian Simon; Henry Stone; Robert Varrin; Werner von Lensa

    2004-06-01

    This report presents the conclusions, observations, and recommendations of the Independent Technology Review Group (ITRG) regarding design features and important technology uncertainties associated with very-high-temperature nuclear system concepts for the Next Generation Nuclear Plant (NGNP). The ITRG performed its reviews during the period November 2003 through April 2004.

  4. 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

  5. Design considerations for multi-core optical fibers in nonlinear switching and mode-locking applications

    CERN Document Server

    Nazemosadat, Elham

    2014-01-01

    We explore the practical challenges which should be addressed when designing a multi-core fiber coupler for nonlinear switching or mode-locking applications. The inevitable geometric imperfections formed in these fiber couplers during the fabrication process affect the performance characteristics of the nonlinear switching device. Fabrication uncertainties are tolerable as long as the changes they impose on the propagation constant of the modes are smaller than the linear coupling between the cores. It is possible to reduce the effect of the propagation constant variations by bringing the cores closer to each other, hence, increasing the coupling. However, higher coupling translates into a higher switching power which may not be desirable in some practical situations. Therefore, fabrication errors limit the minimum achievable switching power in nonlinear couplers.

  6. Spent nuclear fuel project design basis capacity study

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, K.J.

    1996-09-09

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. Alternative configurations, sub-system cycle times, and operating scenarios were tested to identify their impact on total project duration and equipment requirements.

  7. Pulse Design in Solid-State Nuclear Magnetic Resonance

    DEFF Research Database (Denmark)

    Palani, Ravi Shankar

    2017-01-01

    The work presented in this dissertation is centred on the theory of experimental methods in solid-state Nuclear Magnetic Resonance (NMR) spectroscopy, which deals with interaction of electromagnetic radiation with nuclei in a magnetic field and possessing a fundamental quantum mechanical property...

  8. A brief history of design studies on innovative nuclear reactors

    Science.gov (United States)

    Sekimoto, Hiroshi

    2014-09-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  9. A brief history of design studies on innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  10. Exploring the nuclear pasta phase in core-collapse supernova matter.

    Science.gov (United States)

    Pais, Helena; Stone, Jirina R

    2012-10-12

    The core-collapse supernova phenomenon, one of the most explosive events in the Universe, presents a challenge to theoretical astrophysics. Of the large variety of forms of matter present in core-collapse supernova, we focus on the transitional region between homogeneous (uniform) and inhomogeneous (pasta) phases. A three-dimensional, finite temperature Skyrme-Hartree-Fock (3D-SHF)+BCS calculation yields, for the first time fully self-consistently, the critical density and temperature of both the onset of the pasta in inhomogeneous matter, consisting of neutron-rich heavy nuclei and a free neutron and electron gas, and its dissolution to a homogeneous neutron, proton, and electron liquid. We also identify density regions for different pasta formations between the two limits. We employ four different forms of the Skyrme interaction, SkM*, SLy4, NRAPR, and SQMC700 and find subtle variations in the low density and high density transitions into and out of the pasta phase. One new stable pasta shape has been identified, in addition to the classic ones, on the grid of densities and temperatures used in this work. Our results are critically compared to recent calculations of pasta formation in the quantum molecular dynamics approach and Thomas-Fermi and coexisting phase approximations to relativistic mean-field models.

  11. Preliminary issues associated with the next generation nuclear plant intermediate heat exchanger design.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Moisseytsev, A.; Majumdar, S.; Shankar, P. S.; Nuclear Engineering Division

    2007-04-05

    The Next Generation Nuclear Plant (NGNP), which is an advanced high temperature gas reactor (HTGR) concept with emphasis on production of both electricity and hydrogen, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 900-1000 C. In the indirect cycle system, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, nitrogen/helium mixture, or a molten salt. The system concept for the vary high temperature reactor (VHTR) can be a reactor based on the prismatic block of the GT-MHR developed by a consortium led by General Atomics in the U.S. or based on the PBMR design developed by ESKOM of South Africa and British Nuclear Fuels of U.K. This report has made a preliminary assessment on the issues pertaining to the intermediate heat exchanger (IHX) for the NGNP. Two IHX designs namely, shell and tube and compact heat exchangers were considered in the assessment. Printed circuit heat exchanger, among various compact heat exchanger (HX) designs, was selected for the analysis. Irrespective of the design, the material considerations for the construction of the HX are essentially similar, except may be in the fabrication of the units. As a result, we have reviewed in detail the available information on material property data relevant for the construction of HX and made a preliminary assessment of several relevant factors to make a judicious selection of the material for the IHX. The assessment included four primary candidate alloys namely, Alloy 617 (UNS N06617), Alloy 230 (UNS N06230), Alloy 800H (UNS N08810), and Alloy X (UNS N06002) for the IHX. Some of the factors addressed in this report are the tensile, creep, fatigue, creep fatigue, toughness properties for the candidate alloys, thermal aging effects on the mechanical properties, American Society of Mechanical Engineers (ASME) Code compliance

  12. Design data and safety features of commercial nuclear power plants. Vol. I. Docket No. 50-3 through 50-295

    Energy Technology Data Exchange (ETDEWEB)

    Heddleson, F. A.

    1973-12-01

    BS>Design data, safety features, and site characteristics are summarized for thirty-two commercial nuclear power plants in the United States. Six pages of data are presented for each plant consisting of Thermal-Hydraulic and Nuclear Factors, Containment Features, Emergency Core Cooling Systems, Site Features, Circulating Water System Data, and Miscellaneous Factors. An aerial perspective is also presented for each plant. Those covered in this volume are Indian Point No. 1, Docket Number 50-3, and all subsequent plants finishing with Zion, Docket Number 50-295. (auth)

  13. Design data and safety features of commercial nuclear power plants. Vol. I. Docket No. 50-3 through 50-295

    Energy Technology Data Exchange (ETDEWEB)

    Heddleson, F. A.

    1973-12-01

    BS>Design data, safety features, and site characteristics are summarized for thirty-two commercial nuclear power plants in the United States. Six pages of data are presented for each plant consisting of Thermal-Hydraulic and Nuclear Factors, Containment Features, Emergency Core Cooling Systems, Site Features, Circulating Water System Data, and Miscellaneous Factors. An aerial perspective is also presented for each plant. Those covered in this volume are Indian Point No. 1, Docket Number 50-3, and all subsequent plants finishing with Zion, Docket Number 50-295. (auth)

  14. Study on Effect of MOX Fuel Assembly Loaded in Current M310 Reactor Core on Nuclear Design%MOX 燃料组件装入现役 M310堆芯对堆芯核设计的影响研究

    Institute of Scientific and Technical Information of China (English)

    刘晓黎; 宫宇

    2015-01-01

    国际上的 MOX 燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对 MOX 燃料组件的中子学性能进行了分析,对其在我国现役 M310堆芯应用的可行性进行了研究,得到了 M310堆芯由全部使用 UO2燃料组件向使用30%的 MOX 燃料组件过渡的堆芯燃料管理方案,并对使用MOX 燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的 MOX 燃料组件的堆芯可达到与全 UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler 功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为 MOX 燃料在 M310堆芯中应用的进一步研究提供参考。%The MOX fuel technology has been developed and applied in PWR all over the world .In this paper ,the neutronic performance of the MOX fuel assembly was studied , and fuel management scheme of M310 reactor core from all UO2 fuel assemblies to 30%MOX fuel assemblies was given .The results show that the core loaded 30% MOX fuel assemblies can reach the same lifetime as the all UO2 core ,the ability of the control system can meet the requirement of reactivity control ,and the Doppler temperature and power coefficients ,moderator temperature coefficient and the evolutions of Xe and Sm all benefit for the core operation to be more stable .The results of this study prove that the MOX fuel assembly can be used in the M310 reactor core .

  15. Computational Design and Analysis of Core Material of Single-Phase Capacitor Run Induction Motor

    Directory of Open Access Journals (Sweden)

    Gurmeet Singh

    2014-07-01

    Full Text Available A Single-phase induction motor (SPIM has very crucial role in industrial, domestic and commercial sectors. So, the efficient SPIM is a foremost requirement of today's market. For efficient motors, many research methodologies and propositions have been given by researchers in past. Various parameters like as stator/rotor slot variation, size and shape of stator/rotor slots, stator/rotor winding configuration, choice of core material etc. have momentous impact on machine design. Core material influences the motor performance to a degree. Magnetic flux linkage and leakage preliminary depends upon the magnetic properties of core material and air gap. The analysis of effects of core material on the magnetic flux distribution and the performance of induction motor is of immense importance to meet out the desirable performance. An increase in the air gap length will result in the air gap performance characteristics deterioration and decrease in air gap length will lead to serious mechanical balancing concern. So possibility of much variation in air gap beyond the limits on both sides is not feasible. For the optimized performance of the induction motor the core material plays a significant role. Using higher magnetic flux density, reduction on a magnetizing reactance and leakage of flux can be achieved. In this thesis work the analysis of single phase induction motor has been carried out with different core materials. The four models have been simulated using Ansys Maxwell 15.0. Higher flux density selection for same machine dimensions result into huge amount of reduction in iron core losses and thereby improve the efficiency. In this paper 2% higher efficiency has been achieved with Steel_1010 as compared to the machine using conventional D23 material. Out of four models result reflected by the machine using steel_1010 and steel_1008 are found to be better.

  16. Design and testing of coring bits on drilling lunar rock simulant

    Science.gov (United States)

    Li, Peng; Jiang, Shengyuan; Tang, Dewei; Xu, Bo; Ma, Chao; Zhang, Hui; Qin, Hongwei; Deng, Zongquan

    2017-02-01

    Coring bits are widely utilized in the sampling of celestial bodies, and their drilling behaviors directly affect the sampling results and drilling security. This paper introduces a lunar regolith coring bit (LRCB), which is a key component of sampling tools for lunar rock breaking during the lunar soil sampling process. We establish the interaction model between the drill bit and rock at a small cutting depth, and the two main influential parameters (forward and outward rake angles) of LRCB on drilling loads are determined. We perform the parameter screening task of LRCB with the aim to minimize the weight on bit (WOB). We verify the drilling load performances of LRCB after optimization, and the higher penetrations per revolution (PPR) are, the larger drilling loads we gained. Besides, we perform lunar soil drilling simulations to estimate the efficiency on chip conveying and sample coring of LRCB. The results of the simulation and test are basically consistent on coring efficiency, and the chip removal efficiency of LRCB is slightly lower than HIT-H bit from simulation. This work proposes a method for the design of coring bits in subsequent extraterrestrial explorations.

  17. Principles of designing mobile robots for nuclear applications: Some Soviet development projects

    Energy Technology Data Exchange (ETDEWEB)

    Adamov, E.O.; Ivanov, V.G.; Meieran, H.B.

    1990-01-01

    The I.V. Kurchatov Institute of Atomic Energy and the Research and Design Institute of Power Engineering, both designers of nuclear power plant systems and located in Moscow, USSR, have collectively recognized the positive merits of utilizing mobile robots in the nuclear industry. They have given authority to their subsidiary agency CENOTECH to mount an active campaign to program the development of new generations of mobile robots that will support routine and emergency situation operations in the nuclear industry. CENOTECH's rationale for design and performance requirements of mobile robot units to be utilized in the nuclear industry is presented in this paper. A description of design, performance requirements, and operational characteristics of four mobile robots that have been developed at CENOTECH within the past 3 yr is also presented: the 2-tracked KURSOR ; the 4 hybrid-wheeled TELER; the 12-wheeled BUGGY with articulated platforms; and the 2-tracked SADKO.

  18. Design and optimization of 32-core rod/trench assisted square-lattice structured single-mode multi-core fiber.

    Science.gov (United States)

    Xie, Xueqin; Tu, Jiajing; Zhou, Xian; Long, Keping; Saitoh, Kunimasa

    2017-03-06

    We propose and design a kind of heterogeneous rod-assisted and trench-assisted multi-core fiber (Hetero-RA-TA-MCF) with 32 cores arranged in square-lattice structure (SLS), and then we introduce the design method for Hetero-RA-TA-MCF. Simulation results show that the Hetero-RA-TA-32-Core-Fiber achieves average effective area (Aeff) of about 74 μm2, low crosstalk (XT) of about -31 dB/100km, threshold value of bending radius (Rpk) of 7.0 cm, relative core multiplicity factor (RCMF) of 8.74, and cable cutoff wavelength (λcc) of less than 1.53 μm.

  19. Suspended core subwavelength fibers: practical designs for the low-loss terahertz guidance

    CERN Document Server

    Rozé, Mathieu; Mazhorova, Anna; Walther, Markus; Skorobogatiy, Maksim

    2011-01-01

    In this work we report two designs of subwavelength fibers packaged for practical terahertz wave guiding. We describe fabrication, modeling and characterization of microstructured polymer fibers featuring a subwavelength-size core suspended in the middle of a large porous outer cladding. This design allows convenient handling of the subwavelength fibers without distorting their modal profile. Additionally, the air-tight porous cladding serves as a natural enclosure for the fiber core, thus avoiding the need for a bulky external enclosure for humidity-purged atmosphere. Fibers of 5 mm and 3 mm in outer diameters with a 150 \\mu m suspended solid core and a 900 \\mu m suspended porous core respectively, were obtained by utilizing a combination of drilling and stacking techniques. Characterization of the fiber optical properties and the near-field imaging of the guided modes were performed using a terahertz near-field microscopy setup. Near-field imaging of the modal profiles at the fiber output confirmed the effe...

  20. Safeguards Guidance for Designers of Commercial Nuclear Facilities – International Safeguards Requirements for Uranium Enrichment Plants

    Energy Technology Data Exchange (ETDEWEB)

    Philip Casey Durst; Scott DeMuth; Brent McGinnis; Michael Whitaker; James Morgan

    2010-04-01

    For the past two years, the United States National Nuclear Security Administration, Office of International Regimes and Agreements (NA-243), has sponsored the Safeguards-by-Design Project, through which it is hoped new nuclear facilities will be designed and constructed worldwide more amenable to nuclear safeguards. In the course of this project it was recognized that commercial designer/builders of nuclear facilities are not always aware of, or understand, the relevant domestic and international safeguards requirements, especially the latter as implemented by the International Atomic Energy Agency (IAEA). To help commercial designer/builders better understand these requirements, a report was prepared by the Safeguards-by-Design Project Team that articulated and interpreted the international nuclear safeguards requirements for the initial case of uranium enrichment plants. The following paper summarizes the subject report, the specific requirements, where they originate, and the implications for design and construction. It also briefly summarizes the established best design and operating practices that designer/builder/operators have implemented for currently meeting these requirements. In preparing the subject report, it is recognized that the best practices are continually evolving as the designer/builder/operators and IAEA consider even more effective and efficient means for meeting the safeguards requirements and objectives.

  1. A comparison of designer activity using core design situations in the laboratory and practice

    DEFF Research Database (Denmark)

    Cash, Philip; Hicks, Ben J.; Culley, Steve J.

    2013-01-01

    In 2011 one quarter of all articles published in Design Studies and the Journal of Engineering Design used experimental studies. However, there is little work exploring the relationship between laboratory and practice. This paper addresses this by detailing an analysis of designer activity in three...... situations commonly studied by design researchers: information seeking, ideation and design review. This comparison is instantiated through three complementary studies: an observational study of practice and two experimental studies. These reveal a range of similarities and differences that are described...

  2. DESIGN AND CONTROL OF SOAP-FREE HYDROPHILIC-HYDROPHOBIC CORE-SHELL LATEX PARTICLES WITH HIGH CARBOXYL CONTENT IN THE CORE OF THE PARTICLES

    Institute of Scientific and Technical Information of China (English)

    Wen-jiao Ji; Yi-ming Jiang; Bo-tian Li; Wei Deng; Cheng-you Kan

    2012-01-01

    Soap-free hydrophilic-hydrophobic core-shell latex particles with high carboxyl content in the core of the particles were synthesized via the seeded emulsion polymerization using methyl methacrylate (MMA),butyl acrylate (BA),methacrylic acid (MAA),styrene (St) and ethylene glycol dimethacrylate (EGDMA) as monomers,and the influences of MMA content used in the core preparation on polymerization,particle size and morphology were investigated by transmission electron microscopy,dynamic light scattering and conductometric titration.The results showed that the seeded emulsion polymerization could be carried out smoothly using "starved monomer feeding process" when MAA content in the core preparation was equal to or less than 24 wt%,and the encapsulating efficiency of the hydrophilic P(MMA-BA-MAA-EGDMA) core with the hydrophobic PSt shell decreased with the increase in MAA content.When an interlayer of P(MMA-MAA-St) with moderate polarity was inserted between the P(MMA-BA-MAA-EGDMA) core and the PSt shell,well designed soap-free hydrophilic-hydrophobic core-shell latex particles with 24 wt% MAA content in the core preparation were obtained.

  3. The Designing Bus for Nuclear Safety Class Controller

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongil; Lee, Myeongkyun; Yun, Donghwa [PONUTech Co,. Ltd., Research Institute, Ulsan (Korea, Republic of); Ryoo, Kwangki [Hanbat National Univ., Daejeon (Korea, Republic of)

    2013-10-15

    EtherCAT (Ethernet for Control Automation Technology) is based on the IEEE 802.3 standard as one of the communication which is the I/O (Input/Output), sensors and communication function of PLC (Programmable Logic Controller) in industry and factory environment use is increasing. The Nuclear Safety Class Controller implemented by the EtherCAT applied bus can be shown the improving performance of data transmission in the controller.

  4. Spent nuclear fuel project design basis capacity study

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, K.J.

    1998-07-22

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. A base case, reflecting the Fiscal Year 1998 process configuration, is evaluated. Parametric evaluations are also considered, investigating the impact of higher fuel retrieval system productivity and reduced shift operations at the canister storage building on total project duration.

  5. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  6. Design Features of a Core Protection System for an Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Bon Seung; In, Wang Kee; Kim, Keung Koo; Lee, Chung Chan; Zee, Sung Quun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    A system-integrated modular advanced research reactor is under development in the KAERI. Therefore, it is required to design an advanced core protection system for an integral reactor and an online digital core protection system, SCOPS is being developed as a part of plant protection system. SCOPS calculates the minimum CHFR(Critical Heat Flux Ratio) and maximum LPD(Local Power Density) based on the several online measured system parameters, such as the excore detector signal, CEA positions, MCP pump speed, pressure and temperature. Calculated values are compared with predetermined limiting values and the trip signal is generated if necessary. This paper describes the basic design features of SCOPS and several output parameters for a simple test case are presented.

  7. Ultra-large Mode Area Microstructured Core Chalcogenide Fiber Design for Mid-IR Beam Delivery

    CERN Document Server

    Barh, Ajanta; Varshney, R K; Pal, Bishnu P

    2013-01-01

    An all solid large modearea (LMA) chalcogenide based microstructured core optical fiber (MCOF) is designed and proposed for high power handling in the mid IR spectral regime, covering the entire second transparency window of the atmosphere (3 to 5 microns). The core of the proposed specialty fiber is composed of a few rings of high index rods arranged in a pattern of hexagon. Dependence of effective mode area on the pitch and radius of high index rods are studied. Ultra high effective mode area up to 75000 micron square can be achieved over this specific wavelength range while retaining its single mode characteristic. A negligible confinement loss along with a low dispersion slope (near 0.03 ps/km-nm square) and a good beam quality factor (M2 1.17) should make this LMA fiber design attractive for fabrication as a potential candidate suitable for high power, passive applications at the mid IR wavelength regime.

  8. Functional design standard of on-line digital core protection and monitoring systems for SMART

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Kim, Keung Koo; Zee, Sung Qunn [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    The general requirements for the system I/O and the functional design were developed based on the conceptual design of SCOPS and SCOMS for SMART. The reactor trip functions were preliminarily determined to define the design basis events of SCOPS. The sensor requirements for SCOPS and SCOMS were also established. The sensor requirements for SCOPS and SCOMS were also established. The detailed functional design of the SMART digital core protection and monitoring systems will be performed based on the functional design standard in this report. The results of this study will also be useful to determine the reactor trip functions as well as the system and sensor requirements. 3 refs., 2 figs., 5 tabs. (Author)

  9. Design and construction of Nemiscau-1 Dam, the first asphalt core rockfill dam in North America

    Energy Technology Data Exchange (ETDEWEB)

    Alicescu, V.; Tournier, J.P. [Hydro-Quebec, Montreal, PQ (Canada); Vannobel, P. [Societe d' Energie de la Baie James, Montreal, PQ (Canada)

    2008-07-01

    The concept of asphalt as a waterproofing medium inside embankment dams was originally developed in Germany in the 1960s. More than 100 asphalt core rockfill (ACRD) dams have been completed or are under construction. They all have a strong record without any seepage problems or required maintenance. After using the glacial till as waterproofing material for its embankment dams for more than 50 years, Hydro-Quebec is now looking to develop new dam concepts, mainly for the zones where natural waterproofing materials do not exist. In order to do so, the company has decided to design and construct the Nemiscau-1 Dam as a prototype ACRD. This paper presented the detailed design criteria, technical specifications as well as some information concerning the construction of the dam such as asphalt mix design. The given dam site, geology and materials is well suited for a dam with an asphalt core and the chosen core thickness of minimum 400 mm was found to be appropriate, given the small net water head. The main reservoir levels as well as the characteristics of the dam were also listed. Information on the general construction of the dam was provided. It was concluded that the longitudinal profile of the rock excavation and concrete plinth must be optimized, with an optimum balance between the rock excavation, the volume and shape of the concrete plinth and finally, the placement of the asphalt core with the manual method. Several combinations of these 3 elements must be analyzed at the design stage and the most cost effective one should be applied on site. 5 tabs., 7 figs.

  10. Design of Ferrite Core Inductors%铁氧体磁心电感的设计

    Institute of Scientific and Technical Information of China (English)

    毛明; 黄念慈

    2000-01-01

    This paper describes a series of practical equations for the design of ferrite core inductors, and verifies them by measuring and simulating.%整理出一套工程实用的铁氧体磁心电感设计公式,并通过实际测量和计算机仿真对其作了验证。

  11. Design of buffer structure at core nodes in optical burst switching

    Institute of Scientific and Technical Information of China (English)

    LI Lei; ZHANG Min-gde; SUN Xiao-han

    2006-01-01

    Reasonable and effective buffer structures are proposed in core routers /nodes of optical burst switching.Based on the model of burst traffics and their contentions,the basic qualifications for the design of buffer structures are concluded.With these qualifications,buffer and switch integrated structures are proposed;and by conclusion and expansion,the classification rules of buffer structures are also proposed from different angles.The schemes to integrate structures are analyzed and simulated.

  12. Design, synthesis and photochemical properties of the first examples of iminosugar clusters based on fluorescent cores

    Directory of Open Access Journals (Sweden)

    Mathieu L. Lepage

    2015-05-01

    Full Text Available The synthesis and photophysical properties of the first examples of iminosugar clusters based on a BODIPY or a pyrene core are reported. The tri- and tetravalent systems designed as molecular probes and synthesized by way of Cu(I-catalysed azide–alkyne cycloadditions are fluorescent analogues of potent pharmacological chaperones/correctors recently reported in the field of Gaucher disease and cystic fibrosis, two rare genetic diseases caused by protein misfolding.

  13. Optimal core baseline design and observing strategy for probing the astrophysics of reionization with the SKA

    CERN Document Server

    Greig, Bradley; Koopmans, Léon V E

    2015-01-01

    With the first phase of the Square Kilometre Array (SKA1) entering into its final pre-construction phase, we investigate how best to maximise its scientific return. Specifically, we focus on the statistical measurement of the 21 cm power spectrum (PS) from the epoch of reionization (EoR) using the low frequency array, SKA1-low. To facilitate this investigation we use the recently developed MCMC based EoR analysis tool 21CMMC (Greig & Mesinger). In light of the recent 50 per cent cost reduction, we consider several different SKA core baseline designs, changing: (i) the number of antenna stations; (ii) the number of dipoles per station; and also (iii) the distribution of baseline lengths. We find that a design with a reduced number of dipoles per core station (increased field of view and total number of core stations), together with shortened baselines, maximises the recovered EoR signal. With this optimal baseline design, we investigate three observing strategies, analysing the trade-off between lowering t...

  14. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

    Science.gov (United States)

    Adam, Zachary R

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  15. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis

    Science.gov (United States)

    Adam, Zachary R.

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 105-106 years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  16. Framework for Integrating Safety, Operations, Security, and Safeguards in the Design and Operation of Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Darby, John L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Horak, Karl Emanuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaChance, Jeffrey L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Tolk, Keith Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Whitehead, Donnie Wayne [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2007-10-01

    The US is currently on the brink of a nuclear renaissance that will result in near-term construction of new nuclear power plants. In addition, the Department of Energy’s (DOE) ambitious new Global Nuclear Energy Partnership (GNEP) program includes facilities for reprocessing spent nuclear fuel and reactors for transmuting safeguards material. The use of nuclear power and material has inherent safety, security, and safeguards (SSS) concerns that can impact the operation of the facilities. Recent concern over terrorist attacks and nuclear proliferation led to an increased emphasis on security and safeguard issues as well as the more traditional safety emphasis. To meet both domestic and international requirements, nuclear facilities include specific SSS measures that are identified and evaluated through the use of detailed analysis techniques. In the past, these individual assessments have not been integrated, which led to inefficient and costly design and operational requirements. This report provides a framework for a new paradigm where safety, operations, security, and safeguards (SOSS) are integrated into the design and operation of a new facility to decrease cost and increase effectiveness. Although the focus of this framework is on new nuclear facilities, most of the concepts could be applied to any new, high-risk facility.

  17. Natural Nuclear Reactor Oklo and Variation of Fundamental Constants Part 1: Computation of Neutronic of Fresh Core

    CERN Document Server

    Petrov, Yu V; Onegin, M S; Petrov, V Yu; Sakhnovskii, E G; Petrov, Yu.V.

    2006-01-01

    Using a modern methods of reactor physics we have performed the full-scale calculations of the natural reactor Oklo. For reliability we have used the recent version of two Monte Carlo codes: the Russian code MCU REA and world wide known code MCNP (USA). Both codes produce close results. We constructed computer model of zone RZ2 of reactor Oklo which takes into account all details of design and composition. The calculations were performed for the three fresh cores with different uranium contents. Multiplication factors, reactivities and neutron fluxes were calculated. We estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a great difference between reactor spectra and Maxwell's one, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary...

  18. Pygmy and core polarization dipole modes in 206Pb: Connecting nuclear structure to stellar nucleosynthesis

    Science.gov (United States)

    Tonchev, A. P.; Tsoneva, N.; Bhatia, C.; Arnold, C. W.; Goriely, S.; Hammond, S. L.; Kelley, J. H.; Kwan, E.; Lenske, H.; Piekarewicz, J.; Raut, R.; Rusev, G.; Shizuma, T.; Tornow, W.

    2017-10-01

    A high-resolution study of the electromagnetic response of 206Pb below the neutron separation energy is performed using a (γ → ,γ‧) experiment at the HI γ → S facility. Nuclear resonance fluorescence with 100% linearly polarized photon beams is used to measure spins, parities, branching ratios, and decay widths of excited states in 206Pb from 4.9 to 8.1 MeV. The extracted ΣB (E 1) ↑ and ΣB (M 1) ↑ values for the total electric and magnetic dipole strength below the neutron separation energy are 0.9 ± 0.2 e2fm2 and 8.3 ± 2.0 μN2, respectively. These measurements are found to be in very good agreement with the predictions from an energy-density functional (EDF) plus quasiparticle phonon model (QPM). Such a detailed theoretical analysis allows to separate the pygmy dipole resonance from both the tail of the giant dipole resonance and multi-phonon excitations. Combined with earlier photonuclear experiments above the neutron separation energy, one extracts a value for the electric dipole polarizability of 206Pb of αD = 122 ± 10 mb /MeV. When compared to predictions from both the EDF+QPM and accurately calibrated relativistic EDFs, one deduces a range for the neutron-skin thickness of Rskin206 = 0.12- 0.19 fm and a corresponding range for the slope of the symmetry energy of L = 48- 60 MeV. This newly obtained information is also used to estimate the Maxwellian-averaged radiative cross section 205Pb (n , γ)206Pb at 30 keV to be σ = 130 ± 25 mb. The astrophysical impact of this measurement-on both the s-process in stellar nucleosynthesis and on the equation of state of neutron-rich matter-is discussed.

  19. Multiphysics Thermal-Fluid Design Analysis of a Non-Nuclear Tester for Hot-Hydrogen Materials and Component Development

    Science.gov (United States)

    Wang, Ten-See; Foote, John; Litchford, Ron

    2006-01-01

    The objective of this effort is to perform design analyses for a non-nuclear hot-hydrogen materials tester, as a first step towards developing efficient and accurate multiphysics, thermo-fluid computational methodology to predict environments for hypothetical solid-core, nuclear thermal engine thrust chamber design and analysis. The computational methodology is based on a multidimensional, finite-volume, turbulent, chemically reacting, thermally radiating, unstructured-grid, and pressure-based formulation. The multiphysics invoked in this study include hydrogen dissociation kinetics and thermodynamics, turbulent flow, convective, and thermal radiative heat transfers. The goals of the design analyses are to maintain maximum hot-hydrogen jet impingement energy and to minimize chamber wall heating. The results of analyses on three test fixture configurations and the rationale for final selection are presented. The interrogation of physics revealed that reactions of hydrogen dissociation and recombination are highly correlated with local temperature and are necessary for accurate prediction of the hot-hydrogen jet temperature.

  20. Design and performance of a pulse transformer based on Fe-based nanocrystalline core.

    Science.gov (United States)

    Yi, Liu; Xibo, Feng; Lin, Fuchang

    2011-08-01

    A dry-type pulse transformer based on Fe-based nanocrystalline core with a load of 0.88 nF, output voltage of more than 65 kV, and winding ratio of 46 is designed and constructed. The dynamic characteristics of Fe-based nanocrystalline core under the impulse with the pulse width of several microseconds were studied. The pulse width and incremental flux density have an important effect on the pulse permeability, so the pulse permeability is measured under a certain pulse width and incremental flux density. The minimal volume of the toroidal pulse transformer core is determined by the coupling coefficient, the capacitors of the resonant charging circuit, incremental flux density, and pulse permeability. The factors of the charging time, ratio, and energy transmission efficiency in the resonant charging circuit based on magnetic core-type pulse transformer are analyzed. Experimental results of the pulse transformer are in good agreement with the theoretical calculation. When the primary capacitor is 3.17 μF and charge voltage is 1.8 kV, a voltage across the secondary capacitor of 0.88 nF with peak value of 68.5 kV, rise time (10%-90%) of 1.80 μs is obtained.

  1. The Design and Performance of IceCube DeepCore

    CERN Document Server

    ,

    2011-01-01

    The IceCube neutrino observatory in operation at the South Pole, Antarctica, comprises three distinct components: a large buried array for ultrahigh energy neutrino detection, a surface air shower array, and a new buried component called DeepCore. DeepCore was designed to lower the IceCube neutrino energy threshold by over an order of magnitude, to energies as low as about 10 GeV. DeepCore is situated primarily 2100 m below the surface of the icecap at the South Pole, at the bottom center of the existing IceCube array, and began taking physics data in May 2010. Its location takes advantage of the exceptionally clear ice at those depths and allows it to use the surrounding IceCube detector as a highly efficient active veto against the principal background of downward-going muons produced in cosmic-ray air showers. DeepCore has a module density roughly five times higher than that of the standard IceCube array, and uses photomultiplier tubes with a new photocathode featuring a quantum efficiency about 35% higher...

  2. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    Science.gov (United States)

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-12-01

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62149Sm and its dependence on the shift of a resonance position Er (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73⩽ΔEr⩽62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant α. We obtain new, more accurate limits of -4×10-17⩽α·/α⩽3×10-17yr-1. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

  3. First principles design of a core bioenergetic transmembrane electron-transfer protein

    Energy Technology Data Exchange (ETDEWEB)

    Goparaju, Geetha; Fry, Bryan A.; Chobot, Sarah E.; Wiedman, Gregory; Moser, Christopher C.; Leslie Dutton, P.; Discher, Bohdana M.

    2016-05-01

    Here we describe the design, Escherichia coli expression and characterization of a simplified, adaptable and functionally transparent single chain 4-α-helix transmembrane protein frame that binds multiple heme and light activatable porphyrins. Such man-made cofactor-binding oxidoreductases, designed from first principles with minimal reference to natural protein sequences, are known as maquettes. This design is an adaptable frame aiming to uncover core engineering principles governing bioenergetic transmembrane electron-transfer function and recapitulate protein archetypes proposed to represent the origins of photosynthesis. This article is part of a Special Issue entitled Biodesign for Bioenergetics — the design and engineering of electronic transfer cofactors, proteins and protein networks, edited by Ronald L. Koder and J.L. Ross Anderson.

  4. Cost-Optimal Design of a 3-Phase Core Type Transformer by Gradient Search Technique

    Science.gov (United States)

    Basak, R.; Das, A.; Sensarma, A. K.; Sanyal, A. N.

    2014-04-01

    3-phase core type transformers are extensively used as power and distribution transformers in power system and their cost is a sizable proportion of the total system cost. Therefore they should be designed cost-optimally. The design methodology for reaching cost-optimality has been discussed in details by authors like Ramamoorty. It has also been discussed in brief in some of the text-books of electrical design. The paper gives a method for optimizing design, in presence of constraints specified by the customer and the regulatory authorities, through gradient search technique. The starting point has been chosen within the allowable parameter space the steepest decent path has been followed for convergence. The step length has been judiciously chosen and the program has been maneuvered to avoid local minimal points. The method appears to be best as its convergence is quickest amongst different optimizing techniques.

  5. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

    Science.gov (United States)

    2011-03-28

    ... COMMISSION Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs... Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant Designs... Nuclear Power Plants,'' March 2007, Standard Review Plan (SRP), Section 8.3.1 and Sections 9.5.4 through...

  6. 75 FR 5632 - Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using...

    Science.gov (United States)

    2010-02-03

    ... COMMISSION Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using... Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current... for Nuclear Power Plants (LWR Edition),'' June 2007. Background: Emergency diesel generators...

  7. Pygmy and core polarization dipole modes in 206 Pb: Connecting nuclear structure to stellar nucleosynthesis

    Energy Technology Data Exchange (ETDEWEB)

    Tonchev, A. P.; Tsoneva, N.; Bhatia, C.; Arnold, C. W.; Goriely, S.; Hammond, S. L.; Kelley, J. H.; Kwan, E.; Lenske, H.; Piekarewicz, J.; Raut, R.; Rusev, G.; Shizuma, T.; Tornow, W.

    2017-10-01

    A high-resolution study of the electromagnetic response of 206Pb below the neutron separation energy is performed using a (γ→,γ') experiment at the HIγ→S facility. Nuclear resonance fluorescence with 100% linearly polarized photon beams is used to measure spins, parities, branching ratios, and decay widths of excited states in 206Pb from 4.9 to 8.1 MeV. The extracted ΣB(E1)↑ and ΣB(M1)↑ values for the total electric and magnetic dipole strength below the neutron separation energy are 0.9±0.2e2fm2 and 8.3±2.0μ$2\\atop{N}$, respectively. These measurements are found to be in very good agreement with the predictions from an energy-density functional (EDF) plus quasiparticle phonon model (QPM). Such a detailed theoretical analysis allows to separate the pygmy dipole resonance from both the tail of the giant dipole resonance and multi-phonon excitations. Combined with earlier photonuclear experiments above the neutron separation energy, one extracts a value for the electric dipole polarizability of 206Pb of αD=122±10mb/MeV. When compared to predictions from both the EDF+QPM and accurately calibrated relativistic EDFs, one deduces a range for the neutron-skin thickness of R$206\\atop{skin}$=0.12–0.19fm and a corresponding range for the slope of the symmetry energy of L=48–60MeV. This newly obtained information is also used to estimate the Maxwellian-averaged radiative cross section 205Pb(n,γ)Pb206 at 30 keV to be σ=130±25mb. The astrophysical impact of this measurement—on both the s-process in stellar nucleosynthesis and on the equation of state of neutron-rich matter—is discussed.

  8. Design of a homogeneous subcritical nuclear reactor based on thorium with a source of californium 252; Diseno de un reactor nuclear subcritico homogeneo a base de Torio con una fuente de Californio 252

    Energy Technology Data Exchange (ETDEWEB)

    Delgado H, C. E.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: ce_delgado89@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of)

    2015-10-15

    Full text: One of the energy alternatives to fossil fuels which do not produce greenhouse gases is the nuclear energy. One of the drawbacks of this alternative is the generation of radioactive wastes of long half-life and its relation to the generation of nuclear materials to produce weapons of mass destruction. An option to these drawbacks of nuclear energy is to use Thorium as part of the nuclear fuel which it becomes in U{sup 233} when capturing neutrons, that is a fissile material. In this paper Monte Carlo methods were used to design a homogeneous subcritical reactor based on thorium. As neutron reflector graphite was used. The reactor core is homogeneous and is formed of 70% light water as moderator, 12% of enriched uranium UO{sub 2}(NO{sub 3}){sub 4} and 18% of thorium Th(NO{sub 3}){sub 4} as fuel. To start the nuclear fission chain reaction an isotopic source of californium 252 was used with an intensity of 4.6 x 10{sup 7} s{sup -1}. In the design the value of the effective multiplication factor, whose value turned out k{sub eff} <1 was calculated. Also, the neutron spectra at different distances from the source and the total fluence were calculated, as well as the values of the ambient dose equivalent in the periphery of the reactor. (Author)

  9. Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-08-01

    OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.

  10. RF-TSV DESIGN, MODELING AND APPLICATION FOR 3D MULTI-CORE COMPUTER SYSTEMS

    Institute of Scientific and Technical Information of China (English)

    Yu Le; Yang Haigang; Xie Yuanlu

    2012-01-01

    The state-of-the-art multi-core computer systems are based on Very Large Scale three Dimensional (3D) Integrated circuits (VLSI).In order to provide high-speed vertical data transmission in such 3D systems,efficient Through-Silicon Via (TSV) technology is critically important.In this paper,various Radio Frequency (RF) TSV designs and models are proposed.Specifically,the Cu-plug TSV with surrounding ground TSVs is used as the baseline structure.For further improvement,the dielectric coaxial and novel air-gap coaxial TSVs are introduced.Using the empirical parameters of these coaxial TSVs,the simulation results are obtained demonstrating that these coaxial RF-TSVs can provide two-order higher of cut-off frequencies than the Cu-plug TSVs.Based on these new RF-TSV technologies,we propose a novel 3D multi-core computer system as well as new architectures for manipulating the interfaces between RF and baseband circuit.Taking into consideration the scaling down of IC manufacture technologies,predictions for the performance of future generations of circuits are made.With simulation results indicating energy per bit and area per bit being reduced by 7% and 11% respectively,we can conclude that the proposed method is a worthwhile guideline for the design of future multi-core computer ICs.

  11. Helmholtz design for noise transmission attenuation on a chamber core composite cylinder

    Science.gov (United States)

    Li, Deyu; Vipperman, Jeffrey S.

    2002-11-01

    This work explores the feasibility of using Helmholtz resonators to attenuate a subscale ChamberCore cylinder noise transmission. The ChamberCore cylindrical composite is an innovative new sandwich-type structure. It consists of an outer skin, an inner skin, and linking ribs. There are wedge-cross-section chambers along the axis direction between the outer and inner skins. These chambers provide a potential for the acoustic Helmholtz resonator design in order to reduce the noise transmission, which is dominated by the internal acoustic cavity. In this experimental work, the sound transmission behavior of the ChamberCore fairing is investigated and divided into four interesting frequency regions: the stiffness-controlled zone, cavity resonance-controlled zone, coincidence-controlled zone, and mass-controlled zone. It is found that the noise transmission in the low-frequency band is controlled by the structural stiffness and cavity resonances, where the acoustic Helmholtz design method has the potential to improve the noise transmission.

  12. A scalable and low power VLIW DSP core for embedded system design

    Institute of Scientific and Technical Information of China (English)

    Sheraz Anjum; CHEN Jie; HAN Liang; LIN Chuan; ZHANG Xiao-xiao; SU Ye-hua; Chip Cheng

    2008-01-01

    Aims to provide the block architecture of CoStar3400 DSP that is a high performance, low power and scalable VLIW DSP core, it efficiently deployed a variable-length execution set (VLES) execution model which utilizes the maximum parallelism by allowing multiple address generations and data arithmetic logic units to exe-cute multiple instructions in a single clock cycle. The scalability was provided mainly in using more or less num-ber of functional units according to the intended application. Low power support was added by careful architectur-al design techniques such as fine-grain clock gating and activation of only the required number of control signals at each stage of the pipeline. The said features of the core make it a suitable candidate for many SoC configurations,especially for compute intensive applications such as wire-line and wireless communications, including infrastruc-ture and subscriber communications. The embedded system designers can efficiently use the scalability and VLIW features of the core by scaling the number of execution units according to specific needs of the application to effec-tively reduce the power consumption, chip area and time to market the intended final product.

  13. Design of Nuclear Power Plant Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang-ha; Jeong, Yong-hoon; Chang, Soon-heung [KAIST, Daejeon (Korea, Republic of); Lee, Song-kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2007-07-01

    Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability.

  14. Advanced PWR in-core fuel management with optimized gadolinia fuel designs

    Energy Technology Data Exchange (ETDEWEB)

    Berger, H.D.; Neufert, A. [Siemens AG / Power Generation KWU, Nuclear Fuel Cycle, Erlangen (Germany)

    1999-07-01

    Utilities operating LWRs require fuel assemblies and in-core fuel management service, which ensure safe, flexible and cost-effective production of electricity. With the reliability of the fuel having been always the most important requirement, advanced measures to minimize fuel cycle costs are receiving increasing attention in the light of the pressure on costs within the de-regulated power generation markets. The role of in-core fuel management in supporting the goal to minimize fuel cycle costs consists in the development of more demanding core loading strategies, i.e. in the first place more advanced low leakage loading patterns. A prerequisite for this type of loading pattern is the use of an optimized burnable absorber design. Gadolinia as integrated burnable absorber is a very effective means for limiting the critical boron concentration and power peaking factors. Siemens has accumulated extensive experience with Gd-fuel for almost 20 years with e.g. more than 5500 Gd-FA's delivered for PWRs and irradiated up to 65 MWd/kg{sub HM}. Current development efforts for optimizing Gd-fuel are focused on the reduction of the inherent penalties of today's Gd-Fa designs, i.e. reduced average FA enrichment and heavy metal content as well as residual reactivity binding. The most effective way to overcome these drawbacks is the reduction of the Gd{sub 2}O{sub 3} concentration to values of approximately 2 w/o, for which according to recent measurements of the heat conductivity of modern Gd-fuels the reduction of the fissile content in the Gd-rods is no longer necessary. Various feasibility studies have been performed to evaluate the consequences of low-Gd designs for both Siemens PWRs and Non-Siemens PWRs, for which more restrictive boundary conditions with respect to critical boron concentration and peaking factors have to be fulfilled. These studies as well as the first realization of an extended reactor cycle using a low Gd-Fa reload design confirm that the in-core

  15. Applications of fatigue and fracture tolerant design concepts in the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.L.; Marston, T.U.; Tagart, S.W.; Norris, D.M.; Nickell, R.E.

    1982-01-01

    To assure the integrity of nuclear power plant components, fatigue and fracture tolerant design concepts have been incorporated in Sections III and XI of the ASME Code; these contain requirements for nuclear power plant design, construction, and in-service inspection. The methods used in the Code to design against fatigue and brittle fracture are described together with the fracture mechanics based procedure suggested in Sections XI for the evaluation of flaws detected by in-service inspections. Some aspects of the present Code methods that could probably be improved are identified. 19 refs.

  16. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  17. Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng

    2012-01-01

    <正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.

  18. Comparison of Chamfer and Deep Chamfer Preparation Designs on the Fracture Resistance of Zirconia Core Restorations

    Directory of Open Access Journals (Sweden)

    Ezatollah Jalalian

    2011-06-01

    Full Text Available Background and aims. One of the major problems of all-ceramic restorations is their probable fracture under occlusal force. The aim of the present in vitro study was to compare the effect of two marginal designs (chamfer and deep chamfer on the fracture resistance of all-ceramic restorations, CERCON. Materials and methods. This in vitro study was carried out with single-blind experimental technique. One stainless steel die with 50’ chamfer finish line design (0.8 mm deep was prepared using a milling machine. Ten epoxy resin dies were prepared. The same die was retrieved and 50' chamfer was converted into a deep chamfer design (1 mm. Again ten epoxy resin dies were prepared from the deep chamfer die. Zirconia cores with 0.4 mm thickness and 35 µm cement space were fabricated on the epoxy resin dies (10 chamfer and 10 deep chamfer samples. The zirconia cores were cemented on the epoxy resin dies and underwent a fracture test with a universal testing machine and the samples were investigated from the point of view of the origin of the failure. Results. The mean values of fracture resistance for deep chamfer and chamfer samples were 1426.10±182.60 and 991.75±112.00 N, respectively. Student’s t-test revealed statistically significant differences between the groups. Conclusion. The results indicated a relationship between the marginal design of zirconia cores and their fracture resistance. A deep chamfer margin improved the biomechanical performance of posterior single zirconia crown restorations, which might be attributed to greater thickness and rounded internal angles in deep chamfer margins.

  19. Designed armadillo repeat proteins as general peptide-binding scaffolds: consensus design and computational optimization of the hydrophobic core

    DEFF Research Database (Denmark)

    Parmeggiani, Fabio; Pellarin, Riccardo; Larsen, Anders Peter

    2007-01-01

    interactions with peptides or parts of proteins in extended conformation. The conserved binding mode of the peptide in extended form, observed for different targets, makes armadillo repeat proteins attractive candidates for the generation of modular peptide-binding scaffolds. Taking advantage of the large...... number of repeat sequences available, a consensus-based approach combined with a force field-based optimization of the hydrophobic core was used to derive soluble, highly expressed, stable, monomeric designed proteins with improved characteristics compared to natural armadillo proteins. These sequences...

  20. REPORT OF THE WORKSHOP ON NUCLEAR FACILITY DESIGN INFORMATION EXAMINATION AND VERIFICATION FOR SAFEGUARDS

    Energy Technology Data Exchange (ETDEWEB)

    Richard Metcalf; Robert Bean

    2009-10-01

    Executive Summary The International Atomic Energy Agency (IAEA) implements nuclear safeguards and verifies countries are compliant with their international nuclear safeguards agreements. One of the key provisions in the safeguards agreement is the requirement that the country provide nuclear facility design and operating information to the IAEA relevant to safeguarding the facility, and at a very early stage. , This provides the opportunity for the IAEA to verify the safeguards-relevant features of the facility and to periodically ensure that those features have not changed. The national authorities (State System of Accounting for and Control of Nuclear Material - SSAC) provide the design information for all facilities within a country to the IAEA. The design information is conveyed using the IAEA’s Design Information Questionnaire (DIQ) and specifies: (1) Identification of the facility’s general character, purpose, capacity, and location; (2) Description of the facility’s layout and nuclear material form, location, and flow; (3) Description of the features relating to nuclear material accounting, containment, and surveillance; and (4) Description of existing and proposed procedures for nuclear material accounting and control, with identification of nuclear material balance areas. The DIQ is updated as required by written addendum. IAEA safeguards inspectors examine and verify this information in design information examination (DIE) and design information verification (DIV) activities to confirm that the facility has been constructed or is being operated as declared by the facility operator and national authorities, and to develop a suitable safeguards approach. Under the Next Generation Safeguards Initiative (NGSI), the National Nuclear Security Administrations (NNSA) Office of Non-Proliferation and International Security identified the need for more effective and efficient verification of design information by the IAEA for improving international safeguards

  1. Nuclear Safety Functions of ITER Gas Injection System Instrumentation and Control and the Concept Design

    Science.gov (United States)

    Yang, Yu; Maruyama, S.; Fossen, A.; Villers, F.; Kiss, G.; Zhang, Bo; Li, Bo; Jiang, Tao; Huang, Xiangmei

    2016-08-01

    The ITER Gas Injection System (GIS) plays an important role on fueling, wall conditioning and distribution for plasma operation. Besides that, to support the safety function of ITER, GIS needs to implement three nuclear safety Instrumentation and Control (I&C) functions. In this paper, these three functions are introduced with the emphasis on their latest safety classifications. The nuclear I&C design concept is briefly discussed at the end.

  2. Nuclear design manual for generation of cross section and heterogeneous formfunction for CASMO-3/MASTER

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Ho; Cho, Byung Oh; Song, Jae Seong; Lee, Chung Chan

    1996-12-01

    A three-dimensional reactor core simulation code, MASTER, has been developed as a part of the ADONIS project in KAERI. CASMO-3 prepares various two-group cross sections for the constituents of a reactor core such as fuel assembly, radial and axial reflectors, control rod and detector for MASTER. This report includes the standard design procedure for generation of two-group cross sections and heterogeneous formfunction by CASMO-3/FORM for MASTER. (author). 16 refs., 16 tabs., 12 figs.

  3. Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

    2011-02-15

    The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

  4. Ultra-High-Density Molecular Core and Warped Nuclear Disk in the Deep Potential of Radio Lobe Galaxy NGC 3079

    Science.gov (United States)

    Sofue, Y.; Koda, J.; Kohno, K.; Okumura, S. K.; Honma, M.; Kawamura, A.; Irwin, Judith A.

    2001-02-01

    We have performed high-resolution synthesis observations of the 12CO (J=1-0) line emission from the radio lobe edge-on spiral galaxy NGC 3079 using a seven-element millimeter-wave interferometer at the Nobeyama Radio Observatory, which consisted of the 45 m telescope and six-element array. The nuclear molecular disk (NMD) of 750 pc radius is found to be inclined by 20° from the optical disk, and the NMD has spiral arms. An ultra-high-density core (UHC) of molecular gas was found at the nucleus. The gaseous mass of the UHC within 125 pc radius is as large as ~3×108 Msolar, an order of magnitude more massive than that in the same area of the Galactic center, and the mean density is as high as ~3×103H2 cm-3. A position-velocity diagram along the major axis indicates that the rotation curve already starts at a finite velocity exceeding 300 km s-1 from the nucleus. The surface mass density in the central region is estimated to be as high as ~105 Msolar pc-2, producing a very deep gravitational potential. We argue that the very large differential rotation in such a deep potential will keep the UHC gravitationally stable during the current star formation.

  5. The ARIES-RS power core -- Recent development in Li/V designs

    Energy Technology Data Exchange (ETDEWEB)

    Sze, D.K.; Billone, M.C.; Hua, T.Q. [and others

    1997-04-01

    The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirements. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.

  6. Nuclear spins, magnetic moments and quadrupole moments of Cu isotopes from N = 28 to N = 46: probes for core polarization effects

    CERN Document Server

    Vingerhoets, P; Avgoulea, M; Billowes, J; Bissell, M L; Blaum, K; Brown, B A; Cheal, B; De Rydt, M; Forest, D H; Geppert, Ch; Honma, M; Kowalska, M; Kramer, J; Krieger, A; Mane, E; Neugart, R; Neyens, G; Nortershauser, W; Otsuka, T; Schug, M; Stroke, H H; Tungate, G; Yordanov, D T

    2010-01-01

    Measurements of the ground-state nuclear spins, magnetic and quadrupole moments of the copper isotopes from 61Cu up to 75Cu are reported. The experiments were performed at the ISOLDE facility, using the technique of collinear laser spectroscopy. The trend in the magnetic moments between the N=28 and N=50 shell closures is reasonably reproduced by large-scale shell-model calculations starting from a 56Ni core. The quadrupole moments reveal a strong polarization of the underlying Ni core when the neutron shell is opened, which is however strongly reduced at N=40 due to the parity change between the $pf$ and $g$ orbits. No enhanced core polarization is seen beyond N=40. Deviations between measured and calculated moments are attributed to the softness of the 56Ni core and weakening of the Z=28 and N=28 shell gaps.

  7. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  8. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  9. Efficient Design and Analysis of Lightweight Reinforced Core Sandwich and PRSEUS Structures

    Science.gov (United States)

    Bednarcyk, Brett A.; Yarrington, Phillip W.; Lucking, Ryan C.; Collier, Craig S.; Ainsworth, James J.; Toubia, Elias A.

    2012-01-01

    Design, analysis, and sizing methods for two novel structural panel concepts have been developed and incorporated into the HyperSizer Structural Sizing Software. Reinforced Core Sandwich (RCS) panels consist of a foam core with reinforcing composite webs connecting composite facesheets. Boeing s Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) panels use a pultruded unidirectional composite rod to provide axial stiffness along with integrated transverse frames and stitching. Both of these structural concepts are ovencured and have shown great promise applications in lightweight structures, but have suffered from the lack of efficient sizing capabilities similar to those that exist for honeycomb sandwich, foam sandwich, hat stiffened, and other, more traditional concepts. Now, with accurate design methods for RCS and PRSEUS panels available in HyperSizer, these concepts can be traded and used in designs as is done with the more traditional structural concepts. The methods developed to enable sizing of RCS and PRSEUS are outlined, as are results showing the validity and utility of the methods. Applications include several large NASA heavy lift launch vehicle structures.

  10. Analysis and Design of Double-sided Air core Linear Servo Motor with Trapezoidal Permanent Magnets

    DEFF Research Database (Denmark)

    Zhang, Yuqiu; Yang, Zilong; Yu, Minghu

    2011-01-01

    In order to reduce the thrust ripple of linear servo system, a double-sided air core permanent magnet linear servo motor with trapezoidal shape permanent magnets (TDAPMLSM) is proposed in this paper. An analytical model of the motor for predicting the magnetic field in the air-gap at no-load is i......In order to reduce the thrust ripple of linear servo system, a double-sided air core permanent magnet linear servo motor with trapezoidal shape permanent magnets (TDAPMLSM) is proposed in this paper. An analytical model of the motor for predicting the magnetic field in the air-gap at no......-load is introduced. This model is derived based on the equivalent magnetization intensity method, and its accuracy is validated by using the results obtained from the finite-element method. The key dimensions that affect the air-gap magnetic field are analyzed based on the analytical model, and the design...... is optimized by using genetic algorithm. A thrust ripple reduction of 70.6% is achieved by optimization. The proposed analytical model may be used for a quick and reliable design and design optimization of the TDAPMLSM....

  11. Conceptual Design of the Nuclear Electronic Xenon Ion System (NEXIS)

    Science.gov (United States)

    Monheiser, Jeff; Polk, Jay; Randolph, Tom

    2004-01-01

    In support of the NEXIS program, Aerojet-Redmond Operations, with review and input from the JPL and Boeing, has completed the design for a development model (DM) discharge chamber assembly and main discharge cathode assembly. These efforts along with the work by JPL to develop the carbon-carbon-composite ion optics assembly have resulted in a complete ion engine design. The goal of the NEXIS program is to significantly advance the current state of the art by developing an ion engine capable of operating at an input power of 20kW, an Isp of 7500 sec and have a total xenon through put capability of 2000 kg. In this paper we will describe the methodology used to design the discharge chamber and cathode assemblies and describe the resulting final design. Specifics will include the concepts used for the mounting of the ion optics along with the concepts used for the gimbal mounts. In addition, we will present results of a vibrational analysis showing how the engine will respond to a typical Delta IV heavy vibration spectrum.

  12. Development of nuclear design criteria for neutron spallation sources

    Energy Technology Data Exchange (ETDEWEB)

    Sordo, F.; Abanades, A. [E.T.S. Industriales, Madrid Polytechnic University, UPM, J.Gutierrez Abascal, 2 -28006 Madrid (Spain)

    2008-07-01

    Spallation neutron sources allow obtaining high neutronic flux for many scientific and industrial applications. In recent years, several proposals have been made about its use, notably the European Spallation Source (ESS), the Japanese Spallation Source (JSNS) and the projects of Accelerator-Driven Subcritical reactors (ADS), particularly in the framework of EURATOM programs. Given their interest, it seems necessary to establish adequate design basis for guiding the engineering analysis and construction projects of this kind of installations. In this sense, all works done so far seek to obtain particular solutions to a particular design, but there has not been any general development to set up an engineering methodology in this field. In the integral design of a spallation source, all relevant physical processes that may influence its behaviour must be taken into account. Neutronic aspects (emitted neutrons and their spectrum, generation performance..), thermomechanical (energy deposition, cooling conditions, stress distribution..), radiological (spallation waste activity, activation reactions and residual heat) and material properties alteration due to irradiation (atomic displacements and gas generation) must all be considered. After analysing in a systematic manner the different options available in scientific literature, the main objective of this thesis was established as making a significant contribution to determine the limiting factors of the main aspects of spallation sources, its application range and the criteria for choosing optimal materials. To achieve this goal, a series of general simulations have been completed, covering all the relevant physical processes in the neutronic and thermal-mechanical field. Finally, the obtained criteria have been applied to the particular case of the design of the spallation source of subcritical reactors PDX-ADS and XT-ADS. These two designs, developed under the European R and D Framework Program, represent nowadays

  13. Analysis and Design of Double-sided Air core Linear Servo Motor with Trapezoidal Permanent Magnets

    DEFF Research Database (Denmark)

    Zhang, Yuqiu; Yang, Zilong; Yu, Minghu;

    2011-01-01

    In order to reduce the thrust ripple of linear servo system, a double-sided air core permanent magnet linear servo motor with trapezoidal shape permanent magnets (TDAPMLSM) is proposed in this paper. An analytical model of the motor for predicting the magnetic field in the air-gap at no......-load is introduced. This model is derived based on the equivalent magnetization intensity method, and its accuracy is validated by using the results obtained from the finite-element method. The key dimensions that affect the air-gap magnetic field are analyzed based on the analytical model, and the design...

  14. An efficient strategy for designing ambipolar organic semiconductor material: Introducing dehydrogenated phosphorus atoms into pentacene core

    Science.gov (United States)

    Tang, Xiao-Dan

    2017-09-01

    The charge transport properties of phosphapentacene (P-PEN) derivatives were systematically explored by theoretical calculation. The dehydrogenated P-PENs have reasonable frontier molecular orbital energy levels to facilitate both electron and hole injection. The reduced reorganization energies of dehydrogenated P-PENs could be intimately connected to the bonding nature of phosphorus atoms. From the idea of homology modeling, the crystal structure of TIPSE-4P-2p is constructed and fully optimized. Fascinatingly, TIPSE-4P-2p shows the intrinsic property of ambipolar transport in both hopping and band models. Thus, introducing dehydrogenated phosphorus atoms into pentacene core could be an efficient strategy for designing ambipolar material.

  15. IP cores design from specifications to production modeling, verification, optimization, and protection

    CERN Document Server

    Mohamed, Khaled Salah

    2016-01-01

    This book describes the life cycle process of IP cores, from specification to production, including IP modeling, verification, optimization, and protection. Various trade-offs in the design process are discussed, including  those associated with many of the most common memory cores, controller IPs  and system-on-chip (SoC) buses. Readers will also benefit from the author’s practical coverage of new verification methodologies. such as bug localization, UVM, and scan-chain.  A SoC case study is presented to compare traditional verification with the new verification methodologies. ·         Discusses the entire life cycle process of IP cores, from specification to production, including IP modeling, verification, optimization, and protection; ·         Introduce a deep introduction for Verilog for both implementation and verification point of view.  ·         Demonstrates how to use IP in applications such as memory controllers and SoC buses. ·         Describes a new ver...

  16. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2011-09-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

  17. Application of Recommended Design Practices for Conceptual Nuclear Fusion Space Propulsion Systems

    Science.gov (United States)

    Williams, Craig H.

    2004-01-01

    An AIAA Special Project Report was recently produced by AIAA's Nuclear and Future Flight Propulsion Technical Committee and is currently in peer review. The Report provides recommended design practices for conceptual engineering studies of nuclear fusion space propulsion systems. Discussion and recommendations are made on key topics including design reference missions, degree of technological extrapolation and concomitant risk, thoroughness in calculating mass properties (nominal mass properties, weight-growth contingency and propellant margins, and specific impulse), and thoroughness in calculating power generation and usage (power-flow, power contingencies, specific power). The report represents a general consensus of the nuclear fusion space propulsion system conceptual design community and proposes 15 recommendations. This paper expands on the Report by providing specific examples illustrating how to apply each of the recommendations.

  18. A thesis of design air operated value actuator in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. K.; Hwang, J. H.; Kim, Y. B.; Son, K. Ch. [System Design and Development, Daejeon (Korea, Republic of)

    2008-07-01

    AOV used fluid capacity and fluid pressure control in nuclear power plant with heating power plant. AOV structures safely must be secured the reliability and a safety of the atomic power plant. But, AOV where is used from domestic is using the product of the overseas enterprise. The AOV design and maintenance technique is insufficient. Therefore according to ASME designed AOV, the performance test resultant fluid leakage did not occur and AOV design was satisfactory.

  19. Conceptual design report for the ICPP spent nuclear fuel dry storage project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

  20. 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

  1. Elucidating the domain architecture and functions of non-core RAG1: The capacity of a non-core zinc-binding domain to function in nuclear import and nucleic acid binding

    Directory of Open Access Journals (Sweden)

    Zhao Shuying

    2011-05-01

    Full Text Available Abstract Background The repertoire of the antigen-binding receptors originates from the rearrangement of immunoglobulin and T-cell receptor genetic loci in a process known as V(DJ recombination. The initial site-specific DNA cleavage steps of this process are catalyzed by the lymphoid specific proteins RAG1 and RAG2. The majority of studies on RAG1 and RAG2 have focused on the minimal, core regions required for catalytic activity. Though not absolutely required, non-core regions of RAG1 and RAG2 have been shown to influence the efficiency and fidelity of the recombination reaction. Results Using a partial proteolysis approach in combination with bioinformatics analyses, we identified the domain boundaries of a structural domain that is present in the 380-residue N-terminal non-core region of RAG1. We term this domain the Central Non-core Domain (CND; residues 87-217. Conclusions We show how the CND alone, and in combination with other regions of non-core RAG1, functions in nuclear localization, zinc coordination, and interactions with nucleic acid. Together, these results demonstrate the multiple roles that the non-core region can play in the function of the full length protein.

  2. Comparison between direct and indirect cooling core catchers

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jung Soo; Lee, Jong Ho; Bae, Byung Hwan [Korea Hydro and Nuclear Power Co., Ltd., Seoul (Korea, Republic of)

    2012-10-15

    The European nuclear design requirements, which should be satisfied by nuclear reactors in Europe, usually recommend a so called core catcher, which is a molten core ex vessel cooling facility, to manage a severe accident at a nuclear reactor. Two different types of core catcher concepts are compared to determine their abilities to manage severe accidents and cool core melts. The study reveals that direct cooling is better for cooling capacity and is convenient to construct, while indirect cooling is better for the management of a severe accident.

  3. Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

    Science.gov (United States)

    Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

    2015-09-01

    The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with

  4. Security Design of Remote Maintenance Systems for Nuclear Power Plants Based on ISO/IEC 15408

    Science.gov (United States)

    Watabe, Ryosuke; Oi, Tadashi; Endo, Yoshio

    This paper presents a security design of remote maintenance systems for nuclear power plants. Based on ISO/IEC 15408, we list assets to be protected, threats to the assets, security objectives against the threats, and security functional requirements that achieve the security objectives. Also, we show relations between the threats and the security objectives, and relations between the security objectives and the security functional requirements. As a result, we concretize a necessary and sufficient security design of remote maintenance systems for nuclear power plants that can protect the instrumentation and control system against intrusion, impersonation, tapping, obstruction and destruction.

  5. Design concepts for a nuclear digital instrumentation and control system platform

    Energy Technology Data Exchange (ETDEWEB)

    Ou, T. C.; Chen, C. K.; Chen, P. J.; Shyu, S. S.; Lee, C. L. [Institute of Nuclear Energy Research, Atomic Energy Council, No. 1000 Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546 Taiwan (China); Hsieh, S. F., E-mail: tcou@iner.gov.t [Electronics Group, Formosa Plastics Co., 100 Sue-Guan Road, Jen-Wu Hsiang, Kaohsiung County, Taiwan (China)

    2010-10-15

    The objective of this paper is to present the development results of the nuclear instrumentation and control system in Taiwan. As the Taiwan nuclear power plants age, the need to consider upgrading of both their safety and non-safety-related instrumentation and control systems becomes more urgent. Meanwhile, the digital instrumentation and control system that is based on current fast evolving electronic and information technologies are difficult to maintain effectively. Therefore, Institute of Nuclear Energy Research was made a decision to promote the Taiwan Nuclear Instrumentation and Control System project to collaborate with domestic electronic industry to establish self-reliant capabilities on the design, manufacturing, and application of nuclear instrumentation and control systems with newer technology. In the case of safety-related applications like nuclear instrumentation and control, safety-oriented quality control is required. In order to establish a generic qualified digital platform, the world-wide licensing experience should be considered in the licensing process. This paper describes the qualification and certification tools by IEC 61508 for design and development of safety related equipment and explains the basis for many decisions made while performing the digital upgrade. (Author)

  6. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  7. Computer assisted alloy and process design of nuclear structural steels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byeong Joo [Korea Research Institute of Standards and Science, Taejon (Korea, Republic of)

    1997-07-01

    Based on literature research and thermodynamic calculations, it was suggested that toughness of SA508 class 3 steels can be improved by grain refinement using pinning by AIN during forging if alloy contents of Al and N are adjusted. It was also pointed out that the temper embrittlement due to the coarsening of M{sub 2}C carbide may originate from phase transition to the more stable {xi}-carbide. A necessity of experimental works to avoid such a transition by adjustment of alloy composition was claimed. An optimum temperature for the intercirtical heat treatment was derived by thermodynamic= calculation and was found to agree with experimentally derived one. The thermodynamic database and the present calculation scheme can be used as a powerful research tool in further study for design of next generation RPV steels of wide composition range, if combined with the current experimental technology. (Author) 101 refs., 10 tabs., 11 figs.

  8. 10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

    Science.gov (United States)

    2010-01-01

    ... Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FOR NUCLEAR POWER PLANTS Pt. 52, App. N Appendix N to Part 52—Standardization of Nuclear Power Plant... that the applicant wishes to have the application considered under 10 CFR part 52, appendix N, and must...

  9. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  10. Qualification of NEXUS/ANC Nuclear Design System for PWR Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Mayhue, Larry; Milanova, Radka; Huria, Harish; Zhang, Baocheng; Franceschini, Fausto; Ouisloumen, Mohamed [Westinghouse Electric Company, Pittsburgh, PA (United States); Mueller, Erwin; Forslun Guimaraes, Petri [Westinghouse Electric Company, Vaesteraas (Sweden)

    2008-07-01

    NEXUS is a new cross section and nuclear data generation system for core simulators developed by Westinghouse. This system generates once-through, full temperature range nuclear data for both PWRs and BWRs. The system has been implemented for PWRs in the NEXUS/ANC code system. A brief description of the methodology and the codes comprising this system is presented. The qualification for NEXUS/ANC has been completed and a summary of some of the results is presented for 10 plants and 45 cycles of operation. These results include startup data and at-power axial offset performance. Results for low temperature calculations are also presented. The NEXUS/ANC system includes new methodology to cover the operation of AP1000 plants including a new pin power recovery method and a method to capture the effects of control rod depletion. A brief summary of these methods is also presented. (authors)

  11. Experiences in the computerized control rooms design for Nuclear Power Plants; Experiencias en el diseno de salas de control computarizadas para centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Moyano de la Heras, N.; Fernandez Illobre, L.; Valdivia Martin, C.

    2010-07-01

    This paper presents Tecnatom experiences obtained during the control rooms design for the Fuming and Fangjiasham Nuclear Power Plants (CPR type). These are two different locations where two pressurized water reactors, with three loops each one, will be installed.

  12. Nuclear Design Considerations for Z-IFE Chambers

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W R; Schmitt, R C; Abbott, R P; Latkowski, J F; Reyes, S

    2005-02-02

    Z-pinch driven IFE (Z-IFE) requires the design of a repetitive target insertion system that allows coupling of the pulsed power to the target with adequate standoff, and a chamber that can withstand blast and radiation effects from large yield targets. The present strategy for Z-IFE is to use high yield targets ({approx}2-3 GJ/shot), low repetition rate per chamber ({approx}0.1 Hz), and 10 chambers per power plant. In this study, we propose an alternative power plant configuration that uses very high yield targets (20 GJ/shot) in a single chamber operating at 0.1 Hz. A thick-liquid-wall chamber is proposed to absorb the target emission (x-rays, debris and neutrons) and mitigate the blast effects on the chamber wall. The target is attached to the end of a conical shaped Recyclable Transmission Line (RTL) made from a solid coolant (e.g., frozen flibe), or a material that is easily separable from the coolant (e.g., steel). The RTL/target assembly is inserted through a single opening at the top of the chamber for each shot. This study looks at the RTL material choice from a safety and environmental point of view. Materials were assessed according to waste disposal rating (WDR) and contact dose rate (CDR). Neutronics calculations, using the TART2002 Monte Carlo code from Lawrence Livermore National Laboratory (LLNL), were performed for the RTL and Z-IFE chamber, and key results reported here.

  13. User centered design of a digital procedure guidance component for nuclear power plant operation

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Paulo V.R. de; Santos, Isaac L. dos; Oliveira, Mauro V. de; Grecco, Claudio H.S.; Mol, Antonio C. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)]. E-mails: paulov@ien.gov.br; luquetti@ien.gov.br; mvitor@ien.gov.br; grecco@ien.gov.br; mol@ien.gov.br

    2007-07-01

    The use of nuclear power plants to produce electric energy is a safety-critical process where ultimate operational decisions still relies with the control room operators. Thus it is important to provide the best possible decision support through effective supervisory control interfaces. A user centered design approach, based on cognitive task analysis methods, was used to observe the operators training on the nuclear power plant simulator of the Human System Interface Laboratory (LABIHS). We noted deficiencies in the integration between the computerized interfaces and the hardcopy (paper) procedures. An new prototype of digital procedures - the digital procedure component guidance (PCG) - was designed in PowerPoint as an alternative to the current hardcopy procedure manuals. The design improves upon the graphical layout of system information and provides better integration of procedures, automation, and alarm systems. The design was validated by expert opinion and a scenario-based comparison. Future implementation and testing of the redesign is suggested for further validation. (author)

  14. Design Optimization with Geometric Programming for Core Type Large Power Transformers

    Directory of Open Access Journals (Sweden)

    Orosz Tamás

    2014-10-01

    Full Text Available A good transformer design satisfies certain functions and requirements. We can satisfy these requirements by various designs. The aim of the manufacturers is to find the most economic choice within the limitations imposed by the constraint functions, which are the combination of the design parameters resulting in the lowest cost unit. One of the earliest application of the Geometric Programming [GP] is the optimization of power transformers. The GP formalism has two main advantages. First the formalism guarantees that the obtained solution is the global minimum. Second the new solution methods can solve even large-scale GPs extremely efficiently and reliably. The design optimization program seeks a minimum capitalized cost solution by optimally setting the transformer's geometrical and electrical parameters. The transformer's capitalized cost chosen for object function, because it takes into consideration the manufacturing and the operational costs. This paper considers the optimization for three winding, three phase, core-form power transformers. This paper presents the implemented transformer cost optimization model and the optimization results.

  15. Design principles of nuclear receptor signaling: How complex networking improves signal transduction

    NARCIS (Netherlands)

    A.N. Kolodkin (Alexey); F.J. Bruggeman (Frank); N. Plant (Nick); M.J. Moné (Martijn); B.M. Bakker (Barbara); M.J. Campbell (Moray); J.P.T.M. van Leeuwen (Hans); C. Carlberg (Carsten); J.L. Snoep (Jacky); H.V. Westerhoff (Hans)

    2010-01-01

    textabstractThe topology of nuclear receptor (NR) signaling is captured in a systems biological graphical notation. This enables us to identify a number of design aspects of the topology of these networks that might appear unnecessarily complex or even functionally paradoxical. In realistic kinetic

  16. Safety and security aspects in design of digital safety I and C in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Yongjian [University of Applied Sciences Magdeburg-Stendal, Magdeburg (Germany). Inst. of Electrical Engineering; Waedt, Karl [Areva GmbH, Erlangen (Germany). PEAS-G

    2016-05-15

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  17. Design principles of nuclear receptor signaling : how complex networking improves signal transduction

    NARCIS (Netherlands)

    Kolodkin, Alexey N.; Bruggeman, Frank J.; Plant, Nick; Mone, Martijn J.; Bakker, Barbara M.; Campbell, Moray J.; van Leeuwen, Johannes P. T. M.; Carlberg, Carsten; Snoep, Jacky L.; Westerhoff, Hans V.

    2010-01-01

    The topology of nuclear receptor (NR) signaling is captured in a systems biological graphical notation. This enables us to identify a number of 'design' aspects of the topology of these networks that might appear unnecessarily complex or even functionally paradoxical. In realistic kinetic models of

  18. Design principles of nuclear receptor signaling: How complex networking improves signal transduction

    NARCIS (Netherlands)

    A.N. Kolodkin (Alexey); F.J. Bruggeman (Frank); N. Plant (Nick); M.J. Moné (Martijn); B.M. Bakker (Barbara); M.J. Campbell (Moray); J.P.T.M. van Leeuwen (Hans); C. Carlberg (Carsten); J.L. Snoep (Jacky); H.V. Westerhoff (Hans)

    2010-01-01

    textabstractThe topology of nuclear receptor (NR) signaling is captured in a systems biological graphical notation. This enables us to identify a number of design aspects of the topology of these networks that might appear unnecessarily complex or even functionally paradoxical. In realistic kinetic

  19. Preliminary Design of a Manned Nuclear Electric Propulsion Vehicle Using Genetic Algorithms

    Science.gov (United States)

    Irwin, Ryan W.; Tinker, Michael L.

    2005-01-01

    Nuclear electric propulsion (NEP) vehicles will be needed for future manned missions to Mars and beyond. Candidate designs must be identified for further detailed design from a large array of possibilities. Genetic algorithms have proven their utility in conceptual design studies by effectively searching a large design space to pinpoint unique optimal designs. This research combined analysis codes for NEP subsystems with a genetic algorithm. The use of penalty functions with scaling ratios was investigated to increase computational efficiency. Also, the selection of design variables for optimization was considered to reduce computation time without losing beneficial design search space. Finally, trend analysis of a reference mission to the asteroids yielded a group of candidate designs for further analysis.

  20. A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Painter, Chad L.; Geelhood, Ken J.; Wootan, David W.; Meriwether, George H.; Cuta, Judith M.; Adkins, Harold E.; Matson, Dean W.; Abrego, Celestino P.

    2007-12-01

    Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling, core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.