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Sample records for analysis code system

  1. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  2. SINFAC - SYSTEMS IMPROVED NUMERICAL FLUIDS ANALYSIS CODE

    Science.gov (United States)

    Costello, F. A.

    1994-01-01

    The Systems Improved Numerical Fluids Analysis Code, SINFAC, consists of additional routines added to the April 1983 revision of SINDA, a general thermal analyzer program. The purpose of the additional routines is to allow for the modeling of active heat transfer loops. The modeler can simulate the steady-state and pseudo-transient operations of 16 different heat transfer loop components including radiators, evaporators, condensers, mechanical pumps, reservoirs and many types of valves and fittings. In addition, the program contains a property analysis routine that can be used to compute the thermodynamic properties of 20 different refrigerants. SINFAC can simulate the response to transient boundary conditions. SINFAC was first developed as a method for computing the steady-state performance of two phase systems. It was then modified using CNFRWD, SINDA's explicit time-integration scheme, to accommodate transient thermal models. However, SINFAC cannot simulate pressure drops due to time-dependent fluid acceleration, transient boil-out, or transient fill-up, except in the accumulator. SINFAC also requires the user to be familiar with SINDA. The solution procedure used by SINFAC is similar to that which an engineer would use to solve a system manually. The solution to a system requires the determination of all of the outlet conditions of each component such as the flow rate, pressure, and enthalpy. To obtain these values, the user first estimates the inlet conditions to the first component of the system, then computes the outlet conditions from the data supplied by the manufacturer of the first component. The user then estimates the temperature at the outlet of the third component and computes the corresponding flow resistance of the second component. With the flow resistance of the second component, the user computes the conditions down stream, namely the inlet conditions of the third. The computations follow for the rest of the system, back to the first component

  3. System analysis of bar code laser scanner

    Science.gov (United States)

    Wang, Jianpu; Chen, Zhaofeng; Lu, Zukang

    1996-10-01

    This paper focuses on realizing the three important aspects of bar code scanner: generating a high quality scanning light beam, acquiring a fairly even distribution characteristic of light collection, achieving a low signal dynamic range over a large depth of field. To do this, we analyze the spatial distribution and propagation characteristics of scanning laser beam, the vignetting characteristic of optical collection system and their respective optimal design; propose a novel optical automatic gain control method to attain a constant collection over a large working depth.

  4. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

  5. Development of tokamak reactor systems analysis code 'TORSAC'

    International Nuclear Information System (INIS)

    This report describes Tokamak Reactor Systems Analysis Code ''TORSAC'' which has been developed in order to assess the impact of the design choises on reactor systems and to improve tokamak designs in wide parameter range. This computer code has following functions. (1) Systematic sensitivity analysis for a set of given design parameters, (2) Cost calculation of a new reactor concept designed automatically as a result of systematic sensitivity analysis. (author)

  6. Development of tokamak reactor system analysis code NEW-TORSAC

    Science.gov (United States)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  7. Module type plant system dynamics analysis code (MSG-COPD). Code manual

    International Nuclear Information System (INIS)

    MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)

  8. Sub-channel analysis by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

    Alessandro Petruzzi; Anis Bousbia Salah [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Francesco D' Auria [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2005-07-01

    Full text of publication follows: Recent progress in computer technology has increased the possibilities for code calculations in predicting realistically transient scenarios in nuclear power plants. Several attempts have been engaged in order to enlarge the domain for code applications, and to allow best estimate core simulation including interaction effects between neutronics and thermal-hydraulics. In this context, Relap5/Mod3.3 system thermalhydraulic code was used as a sub-channel code for the simulation of the low-pressure boil off experiment No 5002 of Neptun test facility. The experiment constitutes one of the separate effects test (SET) in the OECD/CSNI matrix for thermalhydraulic code validation related to phase separation and vertical flow 'with or without mixture level'. The drying out of the heated elements is expect to occur at very low coolant flow rates, low pressure (about 1.1 bar) and low power level (24.6 kW). The main aim of the activity discussed in the paper is to develop a 'nodalization technology' for accurately modeling the sub-channel grade void distribution problem and in the same way to assess the degree of success in using the Relap5 system code as a sub-channel code for the analysis of local quantities during transients in nuclear reactors. All thermal-hydraulic parameters, such as the collapsed liquid level, critical heat flux time occurrence and heaters surface temperature have been predicted with reasonable accuracy. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. More accurate results have been obtained considering the surface to surface radiation heat transfer model, as well as more cross flow nodes between the test section rods. The overall analysis confirms the possibility of using the Relap5/Mod3.3 system thermal-hydraulic code as sub-channel code to predict the evolution of relevant local quantities measured during 'relevant' experiments

  9. Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations

    International Nuclear Information System (INIS)

    Recently the Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, have developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, two- or three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D. Fluent and RELAP5-3D have strengths that complement one another. CFD codes, such as Fluent, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. On the other hand, RELAP5-3D was developed to analyze the behavior of two-phase systems that could be modeled in one-dimension. Empirical relationships were used where first-principle physics were not well developed. Both Fluent and RELAP5-3D are exemplary in their areas of specialization. The differences between Fluent and RELAP5 fundamentally stem from their field equations. This study focuses on the differences between the momentum equation representations in the two codes (the continuity equation formulations are equivalent for single phase flow). First the differences between the momentum equations are summarized. Next the effect of the differences in the momentum equations are examined by comparing the results obtained using both codes to study the same problem, i.e., fully-developed turbulent pipe flow. Finally, conclusions regarding the significance of the differences are given. (author)

  10. Systemization of burnup sensitivity analysis code (2) (Contract research)

    International Nuclear Information System (INIS)

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

  11. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)

  12. Code Based Analysis for Object-Oriented Systems

    Institute of Scientific and Technical Information of China (English)

    Swapan Bhattacharya; Ananya Kanjilal

    2006-01-01

    The basic features of object-oriented software makes it difficult to apply traditional testing methods in objectoriented systems. Control Flow Graph (CFG) is a well-known model used for identification of independent paths in procedural software. This paper highlights the problem of constructing CFG in object-oriented systems and proposes a new model named Extended Control Flow Graph (ECFG) for code based analysis of Object-Oriented (OO) software. ECFG is a layered CFG where nodes refer to methods rather than statements. A new metrics - Extended Cyclomatic Complexity (E-CC) is developed which is analogous to McCabe's Cyclomatic Complexity (CC) and refers to the number of independent execution paths within the OO software. The different ways in which CFG's of individual methods are connected in an ECFG are presented and formulas for E-CC for these different cases are proposed. Finally we have considered an example in Java and based on its ECFG, applied these cases to arrive at the E-CC of the total system as well as proposed a methodology for calculating the basis set, i.e., the set of independent paths for the OO system that will help in creation of test cases for code testing.

  13. Experimental transport analysis code system in JT-60

    International Nuclear Information System (INIS)

    Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

  14. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

    2015-08-15

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

  15. Analysis of an XADS Target with the System Code TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim; Sanchez Espinoza, Victor H. [Forschungszentrum Karlsruhe GmbH, Institute for Reactor Safety, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Feng, Bo [Massachusetts Institute of Technology, 77 Massachusetts Avenue, NW12-219, Cambridge, MA 02139 (United States)

    2008-07-01

    Accelerator-driven systems (ADS) present an option to reduce the radioactive waste of the nuclear industry. The experimental Accelerator-Driven System (XADS) has been designed to investigate the feasibility of using ADS on an industrial scale to burn minor actinides. The target section lies in the middle of the subcritical core and is bombarded by a proton beam to produce spallation neutrons. The thermal energy produced from this reaction requires a heat removal system for the target section. The target is cooled by liquid lead-bismuth-eutectics (LBE) in the primary system which in turn transfers the heat via a heat exchanger (HX) to the secondary coolant, Diphyl THT (DTHT), a synthetic diathermic fluid. Since this design is still in development, a detailed investigation of the system is necessary to evaluate the behavior during normal and transient operations. Due to the lack of experimental facilities and data for ADS, the analyses are mostly done using thermal hydraulic codes. In addition to evaluating the thermal hydraulics of the XADS, this paper also benchmarks a new code developed by the NRC, TRACE, against other established codes. The events used in this study are beam power switch-on/off transients and a loss of heat sink accident. The obtained results from TRACE were in good agreement with the results of various other codes. (authors)

  16. Automotive Gas Turbine Power System-Performance Analysis Code

    Science.gov (United States)

    Juhasz, Albert J.

    1997-01-01

    An open cycle gas turbine numerical modelling code suitable for thermodynamic performance analysis (i.e. thermal efficiency, specific fuel consumption, cycle state points, working fluid flowrates etc.) of automotive and aircraft powerplant applications has been generated at the NASA Lewis Research Center's Power Technology Division. The use this code can be made available to automotive gas turbine preliminary design efforts, either in its present version, or, assuming that resources can be obtained to incorporate empirical models for component weight and packaging volume, in later version that includes the weight-volume estimator feature. The paper contains a brief discussion of the capabilities of the presently operational version of the code, including a listing of input and output parameters and actual sample output listings.

  17. Phase-Space Analysis of Wavefront Coding Imaging Systems

    Institute of Scientific and Technical Information of China (English)

    YANG Qing-Guo; SUN Jian-Feng; LIU Li-Ren

    2006-01-01

    @@ We explore the use of the Radon-Wigner transform, which is associated with the fractional Fourier transform of the pupil function, for determining the point spread function (PSF) of an incoherent defocused optical system.Then we introduce these phase-space tools to analyse the wavefront coding imaging system. It is shown that the shape of the PSF for such a system is highly invariant to the defocus-related aberrations except for a lateral shift.The optical transfer function of this system is also investigated briefly from a new understanding of ambiguity function.

  18. 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.

  19. Analysis the Performance of Coded WSK-DWDM Transmission System

    Directory of Open Access Journals (Sweden)

    Bobby Barua

    2012-12-01

    Full Text Available Dense Wavelength Division Multiplexing (DWDM is the system with more than eight active wavelengths per fiber. Again high data rates as well as long spans between amplifiers in a chain require high optical power per channel to satisfy the signal to noise ratio (SNR requirements. So the DWDM systems with long repeater-less spans, the simultaneous requirements of high launched power and low dispersion fibers lead to the generation of new waves by four-wave mixing (FWM, which degrades the performance of a multi-channel transmission system. Several methods have been proposed to mitigate the effect of FWM crosstalk. All these works are performed considering only binary WSK scheme. Although M-ary WSK (M>2 schemes have higher spectral efficiency than binary WSK system. Again, the BER performances for M-ary WDM system are not satisfactory with the effect of FWM. Therefore, in this paper we include the effect of FWM on the performance of an M-ary WDM system and try to mitigate the effect by employing the energy efficient convolution code in a normal dispersive fiber as well as in a dispersion shifted fiber (DSF.

  20. Analysis of the KUCA MEU experiments using the ANL code system

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

    1982-01-01

    This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis.

  1. A preliminary uncertainty analysis of phenomenological inputs employed in MAAP code using the SAUNA system

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. H.; Park, S. Y.; Kim, K. R.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Uncertainty analysis is an essential element of safety analysis of nuclear power plants, and especially on the increase as an essential methodology of safety assessment by computer codes. Recently, these efforts have been stepped up to apply the uncertainty methodology in severe accident analysis and PSA Level 2. From this point of view, a statistical sampling-based MAAP-specific platform for a severe accident uncertainty analysis, SAUNA, is being developed in KAERI. Its main purpose is to execute many simulations that are employed for uncertainty analysis. For its efficient implementation, the SAUNA system is composed of three related modules: Firstly, a module for preparing a statistical sampling matrix, secondly, a module for the dynamic linking between code and samples for code simulation, and thirdly, a postprocessing module for further analysis of the code simulation results. The main objective of this paper is to introduce the main functions of the SAUNA system and its example of implementation.

  2. SEACC: the systems engineering and analysis computer code for small wind systems

    Energy Technology Data Exchange (ETDEWEB)

    Tu, P.K.C.; Kertesz, V.

    1983-03-01

    The systems engineering and analysis (SEA) computer program (code) evaluates complete horizontal-axis SWECS performance. Rotor power output as a function of wind speed and energy production at various wind regions are predicted by the code. Efficiencies of components such as gearbox, electric generators, rectifiers, electronic inverters, and batteries can be included in the evaluation process to reflect the complete system performance. Parametric studies can be carried out for blade design characteristics such as airfoil series, taper rate, twist degrees and pitch setting; and for geometry such as rotor radius, hub radius, number of blades, coning angle, rotor rpm, etc. Design tradeoffs can also be performed to optimize system configurations for constant rpm, constant tip speed ratio and rpm-specific rotors. SWECS energy supply as compared to the load demand for each hour of the day and during each session of the year can be assessed by the code if the diurnal wind and load distributions are known. Also available during each run of the code is blade aerodynamic loading information.

  3. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  4. Use of computational fluid dynamics (CFD) codes for safety analysis of nuclear reactor systems, including containment

    International Nuclear Information System (INIS)

    Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The entire collection of papers is provided in this report

  5. HDL code analysis for ASICs in mobile systems

    OpenAIRE

    Wickberg, Fredrik

    2007-01-01

    The complex work of designing new ASICs today and the increasing costs of time to market (TTM) delays are putting high responsibility on the research and development teams to make fault free designs. The main purpose of implementing a static rule checking tool in the design flow today is to find errors and bugs in the hardware definition language (HDL) code as fast and soon as possible. The sooner you find a bug in the design, the shorter the turnaround time becomes, and thereby both time and...

  6. Benchmark analyses of sodium convection in the upper plenum of the MONJU reactor vessel - Comparison between plant system analysis code CERES and CFD code -

    International Nuclear Information System (INIS)

    In the CRP of IAEA, the data of the upper plenum geometry of the prototype FBR“MONJU” and the boundary conditions of the plant trip test were provided by JAEA. A plant system analysis code CERES for FBRs was developed by CRIEPI. To verify the CERES code, analyses had been performed for the system test of the MONJU, the results of which showed good agreement with the test. However, the difficulty of accurately reproducing the temperature variation arising from a complex flow in the upper plenum was identified. By using the general-purpose analysis code STAR-CCM+, detailed analysis in the upper plenum was enabled. Based on comparison between analyses of the CERES and STAR-CCM+ codes, parameters that had to be considered to simulate the flow pattern appropriately for plant system analysis codes were discussed. And, the analysis capability of CERES code with appropriate parameter was able to be confirmed. (author)

  7. EBT reactor systems analysis and cost code: description and users guide (Version 1)

    International Nuclear Information System (INIS)

    An ELMO Bumpy Torus (EBT) reactor systems analysis and cost code that incorporates the most recent advances in EBT physics has been written. The code determines a set of reactors that fall within an allowed operating window determined from the coupling of ring and core plasma properties and the self-consistent treatment of the coupled ring-core stability and power balance requirements. The essential elements of the systems analysis and cost code are described, along with the calculational sequences leading to the specification of the reactor options and their associated costs. The input parameters, the constraints imposed upon them, and the operating range over which the code provides valid results are discussed. A sample problem and the interpretation of the results are also presented

  8. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    Energy Technology Data Exchange (ETDEWEB)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

    2007-03-15

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

  9. SAFIRE: A systems analysis code for ICF [inertial confinement fusion] reactor economics

    International Nuclear Information System (INIS)

    The SAFIRE (Systems Analysis for ICF Reactor Economics) code incorporates analytical models for scaling the cost and performance of several inertial confinement fusion reactor concepts for electric power. The code allows us to vary design parameters (e.g., driver energy, chamber pulse rate, net electric power) and evaluate the resulting change in capital cost of power plant and the busbar cost of electricity. The SAFIRE code can be used to identify the most attractive operating space and to identify those design parameters with the greatest leverage for improving the economics of inertial confinement fusion electric power plants

  10. Analysis of a 12-Finger Rod Drop using RETRAN/MASTER Code System for APR1400

    International Nuclear Information System (INIS)

    The Optimized Power Reactor 1000 (OPR1000) has 4-finger and 12-finger Control Element Assemblies (CEAs). When the 12-finger CEA is dropped, Core Protection Calculator System (CPCS) shuts down the reactor to prevent fuel damage that could occur from the sudden reactor power peaking. By contrast, the improved CPCS of Advanced Power Reactor 1400 (APR1400), which has systems similar to those of the OPR1000, decreases reactor power rapidly using its Reactor Power Cutback System (RPCS) to avoid unwanted reactor trips caused by the CPCS during a 12- finger CEA drop event. RETRAN is a best-estimate code for transient analysis of Non-LOCA. The RETRAN control logic, which includes the function of reducing reactor power during a 12-Finger CEA drop, has been developed for the APR1400. A MATRAN program has also been developed. MATRAN is the interface program for realtime processing to connect RETRAN with MASTER code which is a nuclear analysis and design code. MATRAN supplies adequate feedback reactivities from the MASTER code to RETRAN code. The purpose of this study is to analyze the behavior of a nuclear reactor core and its primary system using conventional RETRAN analysis procedure and MATRAN program analysis procedure during a 12- finger CEA drop. In addition, the axial power distribution and Axial Shape Index (ASI) are produced by the MATRAN program and they are confirmed as within operation limits

  11. Applicability of the SCALE code system to MOX fuel transport systems for criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Toshiaki; Takasugi, Masahiro; Natsume, Toshihiro; Tsuda, Kazuaki

    1996-11-01

    In order to ascertain feasibilities of the SCALE code system for MOX fuel transport systems, criticality analyses were performed for MOX fuel (Pu enrichment; 3.0 wt.%) criticality experiments at JAERI`s TCA and for infinite fuel rod arrays as parameters of Pu enrichment and lattice pitch. The comparison with a combination of the continuous energy Monte Carlo code MCNP and JENDL-3.2 indicated that the SCALE code system with GAM-THERMOS 123-group library can produce feasible results. Though HANSEN-ROACH 16-group library gives poorer results for MOS fuel transport systems, the errors are conservative except for high enriched fuels. (author)

  12. SLSF loop handling system. Volume III. AISC code evaluations and analysis of critical attachments

    International Nuclear Information System (INIS)

    SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions using a linear elastic static equivalent method of stress analysis. Stress computations of Cradle and critical attachments per AISC Code guidelines are presented. HFEF is credited with in-depth review of initial phase of work

  13. CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Nassersharif, B.; Peer, J.S.; DeHart, M.D.

    1987-01-01

    Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators.

  14. Application of data analysis techniques to nuclear reactor systems code to accuracy assessment

    International Nuclear Information System (INIS)

    An automated code assessment program (ACAP) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. This software was developed under subcontract to the United States Nuclear Regulatory Commission for use in its NRS code consolidation efforts. In this paper, background on the topic of NRS accuracy and uncertainty assessment is provided which motivates the development of and defines basic software requirements for ACAP. A survey of data analysis techniques was performed, focusing on the applicability of methods in the construction of NRS code-data comparison measures. The results of this review process, which further defined the scope, user interface and process for using ACAP are also summarized. A description of the software package and several sample applications to NRS data sets are provided. Its functionality and ability to provide objective accuracy assessment figures are demonstrated. (author)

  15. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  16. The stability analysis using two fluids (SAT trademark ) code for boiling flow systems

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.P.; Dykhuizen, R.C.; Su, M.G.; Jain, P. (Arizona State Univ., Tempe, AZ (USA). Dept. of Mechanical and Aerospace Engineering)

    1988-12-01

    This report presents analyses of dynamic instability and frequency response characteristics of boiling flow systems based on an unequal velocity, unequal temperature two-fluid model of such flow. The dynamic instability analyses in the time domain are incorporated into three options of a computer code SAT, viz., DI01 (steady state, or equilibrium point analysis), DI02(linear stability analysis), and DI03 (nonlinear analysis). The frequency response analysis is incorporated into a fourth option FREQ. Results from dynamic instability experiments carried out in a Refrigerant-113 boiling flow rig are also reported as are comparison of these with linear stability analysis predictions. The overall code structures are described in this document, Volume 2. Descriptions of the various subroutines, functions and variables are also included in this volume. 2 refs., 5 figs.

  17. Tokamak Systems Code

    International Nuclear Information System (INIS)

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  18. Development of Three-dimensional Reactor Analysis Code System for Accelerator-Driven System, ADS3D

    International Nuclear Information System (INIS)

    To investigate an Accelerator-Driven System (ADS) with sub-criticality control mechanism such as control rods or burnable poison, the ADS3D code has been developed on MARBLE which is a next generation reactor analysis code system developed by JAEA. In the past neutronics calculation for the ADS, JAEA employed RZ calculation models to realize efficient investigations. However, it was very difficult to model sub-criticality control mechanisms in RZ calculation models. The ADS3D code system is able to calculate the transportation of protons and neutrons, the burn-up calculation and the fuel exchange in three-dimensional calculation models. It means this code system can treat ADS concepts with sub-criticality control mechanism and makes it possible to investigate a new concept of ADS. (author)

  19. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  20. Establishment of joint application system of safety analysis codes between Korea and Vietnam

    International Nuclear Information System (INIS)

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4∼'11.4). 1) On the job training of Vietnam code users(1st training for 4 VAEI staff-3 months. 2nd training for 3 VAEI staff- 3 month), 2) Lecture of nuclear safety analysis (30 hrs basic course and 30 hrs advanced course), 3) Review of safety analysis method (IAEA safety concept and requirements), 4) Collaborative assessment of safety analysis code MARS (13 conceptual problem, 2 separate effect test problem, 1 integral effect test problem), 5) Input deck preparation of standard PWR (Preparation of APR1400 input deck and safety analysis of DBA). VAEI staffs have been familiarized to Korean PWR safety assessment technology through the collaboration assessment work using a computer code developed in Korea. The lectures for Vietnamese research will be contributed to the utilization and cultivation of Korean safety technology. The collaborated assessment works will be used for the establishment of MARS based safety analysis system which is independent from US safety assessment system

  1. Input modelling of PHT system stability analysis for CANFLEX-RU bundle by SOPHT code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    The overall objective of this report is to undertake the stability analysis of primary heat transport (PHT) system for a CANDU reactor to be loaded with CANFLEX-RU bundle which is to provide a vehicle for the economic use of recycled uranium and for the economic provision of additional operating margins in aging CANDU reactors. The modelling report for the input data of SOPHT code is required in order to give the specific and accurate information for flow stability calculation. This report is consisted of the several sections which are described the usage of control cards, calculation methods for input data generation, the comparison of specific input data set for 37-element, CANFLEX-NU and CANFLEX-RU bundles and calculation results for the steady state condition. Those input data set prepared will be used for the flow stability analysis of PHT system of CANDU reactor by SOPHT code. (author). 1 ref., 2 figs., 16 tabs.

  2. PERFORMANCE ANALYSIS OF CHANNEL ESTIMATION FOR LDPC-CODED OFDM SYSTEM IN MULTIPATH FADING CHANNEL

    Institute of Scientific and Technical Information of China (English)

    Zhu Qi; Li Hao; Feng Guangzeng

    2006-01-01

    In this paper, the channel estimation techniques for Orthogonal Frequency Division Multiplexing (OFDM) systems based on pilot arrangement are studied and we apply Low Density Parity Check (LDPC) codes to the system of IEEE 802.16a with OFDM modulation. First investigated is the influence of channel estimation schemes on LDPC-code based OFDM system in static and multipath fading channels. According to the different propagation environments in 802.16a system, a dynamic channel estimation scheme is proposed.A good irregular LDPC code is designed with code rate of 1/2 and code length of 1200. Simulation results show that the performance of LDPC coded OFDM system proposed in this paper is better than that of the convolution Turbo coded OFDM system proposed in IEEE standard 802.16a.

  3. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  4. Combustion chamber analysis code

    Science.gov (United States)

    Przekwas, A. J.; Lai, Y. G.; Krishnan, A.; Avva, R. K.; Giridharan, M. G.

    1993-05-01

    A three-dimensional, time dependent, Favre averaged, finite volume Navier-Stokes code has been developed to model compressible and incompressible flows (with and without chemical reactions) in liquid rocket engines. The code has a non-staggered formulation with generalized body-fitted-coordinates (BFC) capability. Higher order differencing methodologies such as MUSCL and Osher-Chakravarthy schemes are available. Turbulent flows can be modeled using any of the five turbulent models present in the code. A two-phase, two-liquid, Lagrangian spray model has been incorporated into the code. Chemical equilibrium and finite rate reaction models are available to model chemically reacting flows. The discrete ordinate method is used to model effects of thermal radiation. The code has been validated extensively against benchmark experimental data and has been applied to model flows in several propulsion system components of the SSME and the STME.

  5. Calibration curves of a PGNAA system for cement raw material analysis using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Carlos; Salgado, Jose

    1998-12-01

    In large samples, the {gamma}-ray count rate of a prompt gamma neutron activation analysis system is a multi-variable function of the elemental dry composition, density, water content and thickness of the material. The experimental calibration curves require tremendous laboratory work, using a great number of standards with well-known compositions. Although a Monte Carlo simulation study does not avoid the experimental calibration work, it reduces the number of experimental calibration standards. This paper is part of a feasibility study for a PGNAA system for on-line continuous characterisation of cement raw material conveyed on a belt (Oliveira, C., Salgado, J. and Carvalho, F. G. (1997) Optimisation of PGNAA instrument design for cement raw materials using the MCNP code. J. Radioanal. Nucl. Chem. 216(2), 191-198; Oliveira, C., Salgado, J., Goncalves, I. F., Carvalho, F. G. and Leitao, F. (1997a) A Monte Carlo study of the influence of geometry arrangements and structural materials on a PGNAA system performance for cement raw materials analysis. Appl. Radiat. Isot. (accepted); Oliveira, C., Salgado, J. and Leitao, F. (1997b) Density and water content corrections in the gamma count rate of a PGNAA system for cement raw material analysis using the MCNP code. Appl. Radiat. Isot. (accepted).]. It reports on the influence of the density, mass water content and thickness on the calibration curves of the PGNAA system. The MCNP-4A code, running in a Pentium-PC and in a DEC workstation, was used to simulate the PGNAA configuration system.

  6. Calibration curves of a PGNAA system for cement raw material analysis using the MCNP code

    International Nuclear Information System (INIS)

    In large samples, the γ-ray count rate of a prompt gamma neutron activation analysis system is a multi-variable function of the elemental dry composition, density, water content and thickness of the material. The experimental calibration curves require tremendous laboratory work, using a great number of standards with well-known compositions. Although a Monte Carlo simulation study does not avoid the experimental calibration work, it reduces the number of experimental calibration standards. This paper is part of a feasibility study for a PGNAA system for on-line continuous characterisation of cement raw material conveyed on a belt (Oliveira, C., Salgado, J. and Carvalho, F. G. (1997) Optimisation of PGNAA instrument design for cement raw materials using the MCNP code. J. Radioanal. Nucl. Chem. 216(2), 191-198; Oliveira, C., Salgado, J., Goncalves, I. F., Carvalho, F. G. and Leitao, F. (1997a) A Monte Carlo study of the influence of geometry arrangements and structural materials on a PGNAA system performance for cement raw materials analysis. Appl. Radiat. Isot. (accepted); Oliveira, C., Salgado, J. and Leitao, F. (1997b) Density and water content corrections in the gamma count rate of a PGNAA system for cement raw material analysis using the MCNP code. Appl. Radiat. Isot. (accepted).]. It reports on the influence of the density, mass water content and thickness on the calibration curves of the PGNAA system. The MCNP-4A code, running in a Pentium-PC and in a DEC workstation, was used to simulate the PGNAA configuration system

  7. DENINT power plant cost benefit analysis code: Analysis of methane fuelled power plant/district heating system

    International Nuclear Information System (INIS)

    The DENINT power plant cost benefit analysis code takes into consideration, not only power production costs at the generator terminals, but also, in the case of cogeneration, the costs of the fuel supply and heat and power distribution systems which depend greatly on the location of the plant. The code is able to allow comparisons of alternatives with varying annual operation hours, fuel cost increases, and different types of fossil fuels and production systems. For illustrative purposes, this paper examines two methane fired cogeneration plant/district heating alternatives

  8. The stability analysis using two fluids (SAT trademark ) code for boiling flow systems:

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.P.; Dykhuizen, R.C.; Su, M.G.; Jain, P. (Arizona State Univ., Tempe, AZ (USA). Dept. of Mechanical and Aerospace Engineering)

    1988-12-01

    This report presents analyses of dynamic instability and frequency response characteristics of boiling flow systems based on an unequal velocity, unequal temperature two-fluid model of such flow. The dynamic instability analyses in the time domain are incorporated into three options of a computer code SAT, viz., DI01 (steady state, or equilibrium point analysis), DI02 (linear stability analysis), and DI03 (nonlinear analysis). The frequency response analysis is incorporated into a fourth option FREQ. Results of dynamic instability experiments carried out in a Refrigerant-113 boiling flow rig are also reported as are comparison of these with linear stability analysis predictions. This document, Volume 1, provides the theoretical model and computational formulation. The governing conservation equations and constitutive equations of the model are described in Volume 1. Also described are the computational techniques used. 57 refs., 12 figs.

  9. WWER expert system for fuel failure analysis using the RTOP-CA code

    International Nuclear Information System (INIS)

    The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

  10. Using wavefront coding technique as an optical encryption system: reliability analysis and vulnerabilities assessment

    Science.gov (United States)

    Konnik, Mikhail V.

    2012-04-01

    Wavefront coding paradigm can be used not only for compensation of aberrations and depth-of-field improvement but also for an optical encryption. An optical convolution of the image with the PSF occurs when a diffractive optical element (DOE) with a known point spread function (PSF) is placed in the optical path. In this case, an optically encoded image is registered instead of the true image. Decoding of the registered image can be performed using standard digital deconvolution methods. In such class of optical-digital systems, the PSF of the DOE is used as an encryption key. Therefore, a reliability and cryptographic resistance of such an encryption method depends on the size and complexity of the PSF used for optical encoding. This paper gives a preliminary analysis on reliability and possible vulnerabilities of such an encryption method. Experimental results on brute-force attack on the optically encrypted images are presented. Reliability estimation of optical coding based on wavefront coding paradigm is evaluated. An analysis of possible vulnerabilities is provided.

  11. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

    Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

  12. Development of a tritium transport analysis code for the LMFBR system

    Energy Technology Data Exchange (ETDEWEB)

    Iizawa, Katsuyuki; Torii, Tatsuo [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Tsuruga, Fukui (Japan)

    2001-03-01

    A tritium transport analysis code for the LMFBR system, TTT code, has been developed and validated using data from a power rising test conducted at Monju in 1995. The behavior of tritium during future long-term full power operation of Monju has been estimated. The TTT code was created from the tritium and hydrogen transport model devised by R. Kumar and ANL. Actual data from some plants has been used to improve the code. In this study, we used data from Monju to increase the accuracy of the calculated to measured ratio, the C/E ratio. As a result of the study, we were able to: 1. show that the calculated tritium concentration distribution and the change in the primary and secondary sodium, steam and water correlated sufficiently closely with the measured, C/E ratio of 1.1; 2. propose a transport model between sodium and the cover gas system taking into account the mechanisms affecting the partial pressure difference and the isotopic exchange of H and H3; 3. examine the considerable effect of the hydrogen source within the sodium cooling system of Monju on tritium behavior and clarify the characteristics at the initial stage of plant; 4. estimate the tritium transport and distribution for the long-term full power operation of Monju. The tritium release from the core will be 7,400 TBq during 30 years of operation. The primary and secondary cold trap will capture 99% of this and 1% or less will be released to the environment as gaseous radioactive waste from stack and its drainage water from SG; and 5. compare the best fitted tritium source rates from cores in Phenix and Monju and estimate the major release from Monju's helium bond closed type control rods. (author)

  13. Development of a tritium transport analysis code for the LMFBR system

    International Nuclear Information System (INIS)

    A tritium transport analysis code for the LMFBR system, TTT code, has been developed and validated using data from a power rising test conducted at Monju in 1995. The behavior of tritium during future long-term full power operation of Monju has been estimated. The TTT code was created from the tritium and hydrogen transport model devised by R. Kumar and ANL. Actual data from some plants has been used to improve the code. In this study, we used data from Monju to increase the accuracy of the calculated to measured ratio, the C/E ratio. As a result of the study, we were able to: 1. show that the calculated tritium concentration distribution and the change in the primary and secondary sodium, steam and water correlated sufficiently closely with the measured, C/E ratio of 1.1; 2. propose a transport model between sodium and the cover gas system taking into account the mechanisms affecting the partial pressure difference and the isotopic exchange of H and H3; 3. examine the considerable effect of the hydrogen source within the sodium cooling system of Monju on tritium behavior and clarify the characteristics at the initial stage of plant; 4. estimate the tritium transport and distribution for the long-term full power operation of Monju. The tritium release from the core will be 7,400 TBq during 30 years of operation. The primary and secondary cold trap will capture 99% of this and 1% or less will be released to the environment as gaseous radioactive waste from stack and its drainage water from SG; and 5. compare the best fitted tritium source rates from cores in Phenix and Monju and estimate the major release from Monju's helium bond closed type control rods. (author)

  14. Analysis of airborne antenna systems using geometrical theory of diffraction and moment method computer codes

    Science.gov (United States)

    Hartenstein, Richard G., Jr.

    1985-08-01

    Computer codes have been developed to analyze antennas on aircraft and in the presence of scatterers. The purpose of this study is to use these codes to develop accurate computer models of various aircraft and antenna systems. The antenna systems analyzed are a P-3B L-Band antenna, an A-7E UHF relay pod antenna, and traffic advisory antenna system installed on a Bell Long Ranger helicopter. Computer results are compared to measured ones with good agreement. These codes can be used in the design stage of an antenna system to determine the optimum antenna location and save valuable time and costly flight hours.

  15. Application of the coupled code RELAP5-QUABOX/CUBBOX in the system analysis of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bencik, V.; Feretic, D.; Debrecin, N. [Faculty of Electrical Engineering and Computing, Zagreb (Croatia)

    2002-11-01

    Best estimate codes and methods for the realistic simulation of operational transients and accidents are being developed in two directions. First, computer codes with models of the interaction between multidimensional neutron kinetic and NPP dynamic behavior enable realistic simulation of transients characterized by strong coupling between neutronics and thermal-hydraulics as well as of transients that result in asymmetrical spatial core power distribution. Coupled codes consisting of a system thermal-hydraulic code and a multidimensional neutronic code are being developed worldwide in order to accomplish that task. Secondly, development of the qualified plant nodalization and of the models of plant protection and control systems is important for the realistic system analysis of operational transients and accidents. Comparison of the coupled code and point kinetic results is important for the validation of the coupled code and to gain more experience in the use of the coupled code in realistic analyses. In this paper the results of two transients for NPP Krsko using the coupled code RELAP5-QUABOX/CUBBOX (R5QC) and RELAP5 stand alone code are discussed. (orig.)

  16. Coupling CFD code with system code and neutron kinetic code

    International Nuclear Information System (INIS)

    Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent

  17. PERFORMANCE ANALYSIS OF OPTICAL CDMA SYSTEM USING VC CODE FAMILY UNDER VARIOUS OPTICAL PARAMETERS

    Directory of Open Access Journals (Sweden)

    HASSAN YOUSIF AHMED

    2012-06-01

    Full Text Available The intent of this paper is to study the performance of spectral-amplitude coding optical code-division multiple-access (OCDMA systems using Vector Combinatorial (VC code under various optical parameters. This code can be constructed by an algebraic way based on Euclidian vectors for any positive integer number. One of the important properties of this code is that the maximum cross-correlation is always one which means that multi-user interference (MUI and phase induced intensity noise are reduced. Transmitter and receiver structures based on unchirped fiber Bragg grating (FBGs using VC code and taking into account effects of the intensity, shot and thermal noise sources is demonstrated. The impact of the fiber distance effects on bit error rate (BER is reported using a commercial optical systems simulator, virtual photonic instrument, VPITM. The VC code is compared mathematically with reported codes which use similar techniques. We analyzed and characterized the fiber link, received power, BER and channel spacing. The performance and optimization of VC code in SAC-OCDMA system is reported. By comparing the theoretical and simulation results taken from VPITM, we have demonstrated that, for a high number of users, even if data rate is higher, the effective power source is adequate when the VC is used. Also it is found that as the channel spacing width goes from very narrow to wider, the BER decreases, best performance occurs at a spacing bandwidth between 0.8 and 1 nm. We have shown that the SAC system utilizing VC code significantly improves the performance compared with the reported codes.

  18. Analysis of Coded FHSS Systems with Multiple Access Interference over Generalized Fading Channels

    OpenAIRE

    Zummo SalamA

    2008-01-01

    Abstract We study the effect of interference on the performance of coded FHSS systems. This is achieved by modeling the physical channel in these systems as a block fading channel. In the derivation of the bit error probability over Nakagami fading channels, we use the exact statistics of the multiple access interference (MAI) in FHSS systems. Due to the mathematically intractable expression of the Rician distribution, we use the Gaussian approximation to derive the error probability of coded...

  19. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  20. ETR/ITER systems code

    International Nuclear Information System (INIS)

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs

  1. ANALYSIS OF EXISTING AND PROSPECTIVE TECHNICAL CONTROL SYSTEMS OF NUMERIC CODES AUTOMATIC BLOCKING

    Directory of Open Access Journals (Sweden)

    A. M. Beznarytnyy

    2013-09-01

    Full Text Available Purpose. To identify the characteristic features of the engineering control measures system of automatic block of numeric code, identifying their advantages and disadvantages, to analyze the possibility of their use in the problems of diagnosing status of the devices automatic block and setting targets for the development of new diagnostic systems. Methodology. In order to achieve targets the objective theoretical and analytical method and the method of functional analysis have been used. Findings. The analysis of existing and future facilities of the remote control and diagnostics automatic block devices had shown that the existing systems of diagnosis were not sufficiently informative, designed primarily to control the discrete parameters, which in turn did not allow them to construct a decision support subsystem. In developing of new systems of technical diagnostics it was proposed to use the principle of centralized distributed processing of diagnostic data, to include a subsystem support decision-making in to the diagnostics system, it will reduce the amount of work to maintain the devices blocking and reduce recovery time after the occurrence injury. Originality. As a result, the currently existing engineering controls facilities of automatic block can not provide a full assessment of the state distillation alarms and locks. Criteria for the development of new systems of technical diagnostics with increasing amounts of diagnostic information and its automatic analysis were proposed. Practical value. These results of the analysis can be used in practice in order to select the technical control of automatic block devices, as well as the further development of diagnostic systems automatic block that allows for a gradual transition from a planned preventive maintenance service model to the actual state of the monitored devices.

  2. Development of a system of computer codes for severe accident analysis and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Jang, S. H.; Chun, S. W.; Jang, H. S. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1993-01-15

    As a continuing study for the development of a system of computer codes to analyze severe accidents which had been performed last year, major focuses were on the aspect of application of the developed code systems. As the first step, two most commonly used code packages other than STCP, i.e., MELCOR of NRC and MAAP of IDCOR were reviewed to compare the models that they used. Next, important heat transfer phenomena were surveyed as severe accident progressed. Particularly, debris bed coolability and molten core-concrete interaction were selected as sample models, and they were studied extensively. The recent theoretical works and experiments for these phenomena were surveyed, and also the relevant models adopted by major code packages were compared and assessed. Based on the results obtained in this study, it is expected to be able to take into account these phenomenological uncertainties when one uses the severe accident code packages for probabilistic safety assessments or accident management programs.

  3. Accumulative Landings System Code Tables

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Code Tables Used In Landings System. These tables assign meanings to the codes that appear in the data tables. Code tables exist for species, gear, state, county,...

  4. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  5. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  6. Electrical utility generating system reliability analysis code, SYSREL. Social cost studies program

    Energy Technology Data Exchange (ETDEWEB)

    Hub, K.; Conley, L.; Buehring, W.; Rowland, B.; Stephenson, M.

    1975-09-01

    The system reliability code, SYSREL, is a system planning tool that can be used to assess the reliability and economic performance of alternative expansion patterns of electric utility generation systems. Given input information such as capacity, forced-outage rate, number of weeks of annual scheduled maintenance, and economic data for individual units along with the expected load characteristics, the code produces estimates of the mean time between system failures, required reserve capacity to meet a specified system-failure-frequency criterion, expected energy generation from each unit, and system energy cost. The categories of calculations performed by the code are maintenance scheduling, reliability, capacity requirement, energy production allocation, and energy cost. The code is designed to examine alternative generating units and system expansion patterns based on the constraints and general economic conditions imposed by the investigator. The computer running time to execute a study is short and many system alternatives can be examined at a relatively low cost. The report contains a technical description of the code, list of input data requirements, program listing, sample execution, and parameter studies. (auth)

  7. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

    This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

  9. Analysis of Coded FHSS Systems with Multiple Access Interference over Generalized Fading Channels

    Directory of Open Access Journals (Sweden)

    Zummo SalamA

    2008-01-01

    Full Text Available Abstract We study the effect of interference on the performance of coded FHSS systems. This is achieved by modeling the physical channel in these systems as a block fading channel. In the derivation of the bit error probability over Nakagami fading channels, we use the exact statistics of the multiple access interference (MAI in FHSS systems. Due to the mathematically intractable expression of the Rician distribution, we use the Gaussian approximation to derive the error probability of coded FHSS over Rician fading channel. The effect of pilot-aided channel estimation is studied for Rician fading channels using the Gaussian approximation. From this, the optimal hopping rate in coded FHSS is approximated. Results show that the performance loss due to interference increases as the hopping rate decreases.

  10. Analysis of Coded FHSS Systems with Multiple Access Interference over Generalized Fading Channels

    Directory of Open Access Journals (Sweden)

    Salam A. Zummo

    2009-02-01

    Full Text Available We study the effect of interference on the performance of coded FHSS systems. This is achieved by modeling the physical channel in these systems as a block fading channel. In the derivation of the bit error probability over Nakagami fading channels, we use the exact statistics of the multiple access interference (MAI in FHSS systems. Due to the mathematically intractable expression of the Rician distribution, we use the Gaussian approximation to derive the error probability of coded FHSS over Rician fading channel. The effect of pilot-aided channel estimation is studied for Rician fading channels using the Gaussian approximation. From this, the optimal hopping rate in coded FHSS is approximated. Results show that the performance loss due to interference increases as the hopping rate decreases.

  11. Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

  12. Performance analysis and code recognition for dual N-ary orthogonal hybrid modulation systems

    Institute of Scientific and Technical Information of China (English)

    Qiao Xiaoqiang; Zhao Hangsheng; Cai Yueming

    2008-01-01

    A dual N-ary orthogonal hybrid modulation system is introduced in this paper, which can increase the data rate greatly compared with conventional N-ary orthogonal spread spectrum system, so it can be used for high rate data communication. Then, three code recognition algorithms are presented for dual N-ary orthogonal hybrid modulation system and the analytic bit error rate (BER) performance of the system in additive white Gaussian noise (AWGN) and flat Rayleigh fading channel is derived. Finally, the computer simulation of the system with three code recognition algorithms is performed, which shows that the simplified maximum a posteriori (MAP) algorithm is the best for the system with a compromise between the performance and the complexity.

  13. Optical transfer function analysis of circular-pupil wavefront coding systems with separable phase masks

    Institute of Scientific and Technical Information of China (English)

    Zhao Ting-Yu; Liu Qin-Xiao; Yu Fei-Hong

    2012-01-01

    This paper proposes a simple method to achieve the optical transfer function of a circular pupil wavefront coding system with a separable phase mask in Cartesian coordinates.Based on the stationary phase method,the optical transfer function of the circular pupil system can be easily obtained from the optical transfer function of the rectangular pupil system by modifying the cut-off frequency and the on-axial modulation transfer function.Finally,a system with a cubic phase mask is used as an example to illustrate the way to achieve the optical transfer function of the circular pupil system from the rectangular pupil system.

  14. Development of core thermal hydraulic analysis methodology using multichannel code system

    International Nuclear Information System (INIS)

    A multi-channel core analysis model using a subchannel code TORC is developed to improve the thermal margin, and is assessed and compared with the existing single-channel analysis model. To apply the TORC code to the w-type reactor core, a hot subchannel DNBR analysis model is developed using the lumping technology. In addition, the sensitivity of TORC to various models and input parameters are carried out to appreciate the code characteristics. The developed core analysis model is applied to the evaluation of the thermal margin for 17 x 17 KOFA loaded core. For this calculation, the KRB1 CHF correlation is developed on the basis of w and Siemens bundle CHF data, and the DNB design limit is established using the STDP method. From the result of the steady-state and transient analysis of the 17 x 17 KOFA loaded core, it is found that the extra 10% DNBR margin can be obtained compared with the existing single-channel analysis methodology. (Author) 65 figs., 12 tabs

  15. Users manual for the FORSS sensitivity and uncertainty analysis code system

    International Nuclear Information System (INIS)

    FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions and associated uncertainties. This report describes the computing environment and the modules currently used to implement FORSS Sensitivity and Uncertainty Methodology

  16. Users manual for the FORSS sensitivity and uncertainty analysis code system

    Energy Technology Data Exchange (ETDEWEB)

    Lucius, J.L.; Weisbin, C.R.; Marable, J.H.; Drischler, J.D.; Wright, R.Q.; White, J.E.

    1981-01-01

    FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions and associated uncertainties. This report describes the computing environment and the modules currently used to implement FORSS Sensitivity and Uncertainty Methodology.

  17. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  18. Sub-channel analysis by RELAP5 system code of boil-off experiment (Test 5002) with NEPTUN facility

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, A. [Pennsylvania State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, Pennsylvania (United States)]. E-mail: axp46@psu.edu; Bousbia Salah, A.; D' Auria, F. [Univ. of Pisa, Dipartimento di Ingegneria Meccanica, Nucleare d della Produzione, Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; f.dauria@ing.unipi.it

    2004-07-01

    This paper presents the results of RELAP5/Mod3.2 system thermalhydraulic code using the sub-channel analysis approach in predicting the NEPTUN separate effect boil off experiments. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the NEPTUN low pressure test N{sup o}5002 has been considered. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory and demonstrate, as well, the reasonable success of the 'sub-channel analysis' approach adopted in the present context for a system thermalhydraulic code. (author)

  19. Development of a system analysis code, SSC-K, for inherent safety evaluation of the Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development. This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram

  20. Development of realistic thermal-hydraulic system analysis codes ; development of thermal hydraulic test requirements for multidimensional flow modeling

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)

    2002-03-01

    This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)

  1. Development of system of computer codes for severe accident analysis and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Jang, H. S.; Jeon, M. H.; Cho, N. J. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1992-01-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts.

  2. Using finite mixture models in thermal-hydraulics system code uncertainty analysis

    International Nuclear Information System (INIS)

    Highlights: • Best estimate codes simulation needs uncertainty quantification. • The output variables can present multimodal probability distributions. • The analysis of multimodal distribution is performed using finite mixture models. • Two methods to reconstruct output variable probability distribution are used. -- Abstract: Nuclear Power Plant safety analysis is mainly based on the use of best estimate (BE) codes that predict the plant behavior under normal or accidental conditions. As the BE codes introduce uncertainties due to uncertainty in input parameters and modeling, it is necessary to perform uncertainty assessment (UA), and eventually sensitivity analysis (SA), of the results obtained. These analyses are part of the appropriate treatment of uncertainties imposed by current regulation based on the adoption of the best estimate plus uncertainty (BEPU) approach. The most popular approach for uncertainty assessment, based on Wilks’ method, obtains a tolerance/confidence interval, but it does not completely characterize the output variable behavior, which is required for an extended UA and SA. However, the development of standard UA and SA impose high computational cost due to the large number of simulations needed. In order to obtain more information about the output variable and, at the same time, to keep computational cost as low as possible, there has been a recent shift toward developing metamodels (model of model), or surrogate models, that approximate or emulate complex computer codes. In this way, there exist different techniques to reconstruct the probability distribution using the information provided by a sample of values as, for example, the finite mixture models. In this paper, the Expectation Maximization and the k-means algorithms are used to obtain a finite mixture model that reconstructs the output variable probability distribution from data obtained with RELAP-5 simulations. Both methodologies have been applied to a separated

  3. Parallel processing of structural integrity analysis codes

    International Nuclear Information System (INIS)

    Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab

  4. Code C# for chaos analysis of relativistic many-body systems with reactions

    Science.gov (United States)

    Grossu, I. V.; Besliu, C.; Jipa, Al.; Stan, E.; Esanu, T.; Felea, D.; Bordeianu, C. C.

    2012-04-01

    In this work we present a reaction module for “Chaos Many-Body Engine” (Grossu et al., 2010 [1]). Following our goal of creating a customizable, object oriented code library, the list of all possible reactions, including the corresponding properties (particle types, probability, cross section, particle lifetime, etc.), could be supplied as parameter, using a specific XML input file. Inspired by the Poincaré section, we propose also the “Clusterization Map”, as a new intuitive analysis method of many-body systems. For exemplification, we implemented a numerical toy-model for nuclear relativistic collisions at 4.5 A GeV/c (the SKM200 Collaboration). An encouraging agreement with experimental data was obtained for momentum, energy, rapidity, and angular π distributions. Catalogue identifier: AEGH_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGH_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 184 628 No. of bytes in distributed program, including test data, etc.: 7 905 425 Distribution format: tar.gz Programming language: Visual C#.NET 2005 Computer: PC Operating system: Net Framework 2.0 running on MS Windows Has the code been vectorized or parallelized?: Each many-body system is simulated on a separate execution thread. One processor used for each many-body system. RAM: 128 Megabytes Classification: 6.2, 6.5 Catalogue identifier of previous version: AEGH_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181 (2010) 1464 External routines: Net Framework 2.0 Library Does the new version supersede the previous version?: Yes Nature of problem: Chaos analysis of three-dimensional, relativistic many-body systems with reactions. Solution method: Second order Runge-Kutta algorithm for simulating relativistic many-body systems with reactions

  5. Development of heat and mass balance analysis code in out-of-pile hydrogen production system for HTTR heat utilization system (contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Inaba, Yoshitomo; Inagaki, Yoshiyuki; Hayashi, Koji; Suyama, Kazumasa [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-03-01

    A heat and mass balance analysis code has been developed to examine test conditions, to investigate transient behavior etc. in the out-of-pile hydrogen production system for the HTTR heat utilization system. The code can analyze temperature, mass and pressure profiles of helium and process gases and behavior of the control system under both static state (case of steady operation) and dynamic state (case of transient operation). This report describes analytical methods, basic equations and constitution of the code, and how to make of the input data, estimate of the analytical results and so on. (author)

  6. Tunable wavefront coded imaging system based on detachable phase mask: Mathematical analysis, optimization and underlying applications

    Science.gov (United States)

    Zhao, Hui; Wei, Jingxuan

    2014-09-01

    The key to the concept of tunable wavefront coding lies in detachable phase masks. Ojeda-Castaneda et al. (Progress in Electronics Research Symposium Proceedings, Cambridge, USA, July 5-8, 2010) described a typical design in which two components with cosinusoidal phase variation operate together to make defocus sensitivity tunable. The present study proposes an improved design and makes three contributions: (1) A mathematical derivation based on the stationary phase method explains why the detachable phase mask of Ojeda-Castaneda et al. tunes the defocus sensitivity. (2) The mathematical derivations show that the effective bandwidth wavefront coded imaging system is also tunable by making each component of the detachable phase mask move asymmetrically. An improved Fisher information-based optimization procedure was also designed to ascertain the optimal mask parameters corresponding to specific bandwidth. (3) Possible applications of the tunable bandwidth are demonstrated by simulated imaging.

  7. Elements of algebraic coding systems

    CERN Document Server

    Cardoso da Rocha, Jr, Valdemar

    2014-01-01

    Elements of Algebraic Coding Systems is an introductory textto algebraic coding theory. In the first chapter, you'll gain insideknowledge of coding fundamentals, which is essential for a deeperunderstanding of state-of-the-art coding systems.This book is a quick reference for those who are unfamiliar withthis topic, as well as for use with specific applications such as cryptographyand communication. Linear error-correcting block codesthrough elementary principles span eleven chapters of the text.Cyclic codes, some finite field algebra, Goppa codes, algebraic decodingalgorithms, and applications in public-key cryptography andsecret-key cryptography are discussed, including problems and solutionsat the end of each chapter. Three appendices cover the Gilbertbound and some related derivations, a derivation of the Mac-Williams' identities based on the probability of undetected error,and two important tools for algebraic decoding-namely, the finitefield Fourier transform and the Euclidean algorithm for polynomials.

  8. Integrated verification test of Severe Accident Analysis Code SAMPSON in super Simulation 'IMPACT' system

    Energy Technology Data Exchange (ETDEWEB)

    Ujita, Hiroshi; Naitoh, Masanori [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan); Karasawa, Hidetoshi; Miyagi, Kazumi

    1999-07-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology', project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analysed and phenomena occurred in scenarios can be simulated quantitatively reasonably considering the physical models used for the situation. (author)

  9. Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

    International Nuclear Information System (INIS)

    This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author)

  10. A New Performance Analysis Method of Optical Code Division Multiple Access Systems with An Optical Hard-Limiter

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A new performance analysis method of Optical Code Division Multiple Access (OCDMA) systems with an optical hard-limiter is studied. The bit error probability of the OCDMA system is derived, and the numerical results of the system with and without an ideal optical hard-limiter are analyzed respectively. The results show that although the derived expression is different from the one derived by J A Salehi[1], the numerical results are the same as those analyzed by J A Salehi, and the numerical result can be easily achieved in this expression.

  11. Performance analysis of electronic structure codes on HPC systems: a case study of SIESTA.

    Directory of Open Access Journals (Sweden)

    Fabiano Corsetti

    Full Text Available We report on scaling and timing tests of the SIESTA electronic structure code for ab initio molecular dynamics simulations using density-functional theory. The tests are performed on six large-scale supercomputers belonging to the PRACE Tier-0 network with four different architectures: Cray XE6, IBM BlueGene/Q, BullX, and IBM iDataPlex. We employ a systematic strategy for simultaneously testing weak and strong scaling, and propose a measure which is independent of the range of number of cores on which the tests are performed to quantify strong scaling efficiency as a function of simulation size. We find an increase in efficiency with simulation size for all machines, with a qualitatively different curve depending on the supercomputer topology, and discuss the connection of this functional form with weak scaling behaviour. We also analyze the absolute timings obtained in our tests, showing the range of system sizes and cores favourable for different machines. Our results can be employed as a guide both for running SIESTA on parallel architectures, and for executing similar scaling tests of other electronic structure codes.

  12. Flow Analysis of Code Customizations

    DEFF Research Database (Denmark)

    Hessellund, Anders; Sestoft, Peter

    2008-01-01

    Inconsistency between metadata and code customizations is a major concern in modern, configurable enterprise systems. The increasing reliance on metadata, in the form of XML files, and code customizations, in the form of Java files, has led to a hybrid development platform. The expected consistency...

  13. A new code for the design and analysis of the heliostat field layout for power tower system

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Xiudong; Lu, Zhenwu; Yu, Weixing [Changchun Institute of Optics, Fine Mechanics and Physics of Chinese Academy of Sciences, Changchun 130033 (China); Wang, Zhifeng [The Key Laboratory of Solar Thermal Energy and Photovoltaic system, Institute of Electrical Engineering, Chinese Academy of Sciences, Beijing 100190 (China)

    2010-04-15

    A new code for the design and analysis of the heliostat field layout for power tower system is developed. In the new code, a new method for the heliostat field layout is proposed based on the edge ray principle of nonimaging optics. The heliostat field boundary is constrained by the tower height, the receiver tilt angle and size and the heliostat efficiency factor which is the product of the annual cosine efficiency and the annual atmospheric transmission efficiency. With the new method, the heliostat can be placed with a higher efficiency and a faster response speed of the design and optimization can be obtained. A new module for the analysis of the aspherical heliostat is created in the new code. A new toroidal heliostat field is designed and analyzed by using the new code. Compared with the spherical heliostat, the solar image radius of the field is reduced by about 30% by using the toroidal heliostat if the mirror shape and the tracking are ideal. In addition, to maximize the utilization of land, suitable crops can be considered to be planted under heliostats. To evaluate the feasibility of the crop growth, a method for calculating the annual distribution of sunshine duration on the land surface is developed as well. (author)

  14. An Introduction to Thermodynamic Performance Analysis of Aircraft Gas Turbine Engine Cycles Using the Numerical Propulsion System Simulation Code

    Science.gov (United States)

    Jones, Scott M.

    2007-01-01

    This document is intended as an introduction to the analysis of gas turbine engine cycles using the Numerical Propulsion System Simulation (NPSS) code. It is assumed that the analyst has a firm understanding of fluid flow, gas dynamics, thermodynamics, and turbomachinery theory. The purpose of this paper is to provide for the novice the information necessary to begin cycle analysis using NPSS. This paper and the annotated example serve as a starting point and by no means cover the entire range of information and experience necessary for engine performance simulation. NPSS syntax is presented but for a more detailed explanation of the code the user is referred to the NPSS User Guide and Reference document (ref. 1).

  15. User's guide for the JULIET module of the FORSS sensitivity and uncertainty analysis code system

    International Nuclear Information System (INIS)

    JULIET is the FORSS module that calculates generalized sources, responses (e.g., criticality, reaction rate ratios, reactivity worths), normalization parameters and sensitivity coefficients. JULIET is organized into execution paths which are in effect submodules. This design permits a problem to be segmented for solution at the user's discretion (i.e., multiple entry points). JULIET normally operates with fluxes generated by the FORSS version of ANISM; however, the execution path concept permits interaction with other neutronics codes such as DOT and VENTURE. The proposed CCCC file MATXS is the cross-section data base for JULIET permitting the calcuation of sensitivity coefficients with respect to partial cross sections. The sensitivity coefficients calculated by JULIET are placed in the proposed CCCC file SENPRO where they may be accessed by other modules in the FORSS system or transmitted to other installations. 1 figure

  16. The stability analysis using two fluids (SAT) code for boiling flow systems: Volume 4, Experiments and model validation

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.P.; Dykhuizen, R.C.; Su, M.G.; Jain, P.

    1988-12-01

    This report presents analyses of dynamic instability and frequency response characteristics of boiling flow systems based on an unequal velocity, unequal temperature two-fluid model of such flow. The dynamic instability analyses in the time domain are incorporated into three options of a computer code SAT (steady state, or equilibrium point analyses; linear stability analysis; and nonlinear analysis). The frequency response analysis is incorporated into a fourth option FREQ. Results from dynamic instability experiments carried out in a Refrigerant-113 boiling flow rig are also reported as are comparison of these with linear stability analysis predictions. Descriptions of the model, the computational techniques, the computer codes, the experiments and model validation are divided into the following volumes: Volume 1, theoretical model and computational formulation; Volume 2, coding description; Volume 3, user's manual; and Volume 4, experiments and model validation. Instability experiments run in our Refrigerant-113 boiling flow facility are described in this document. Results from these experiments are compared with predictions of the theoretical model. Instability experiment data from two other facilities and frequency response results from one are compared with theoretical model predictions also. 19 refs., 41 figs.

  17. Using the RELAP5-3D advanced systems analysis code with commercial and advanced CFD software

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. A demonstration calculation is presented. Finally, future tasks and plans are outlined. (author)

  18. Automated Facial Action Coding System for dynamic analysis of facial expressions in neuropsychiatric disorders.

    Science.gov (United States)

    Hamm, Jihun; Kohler, Christian G; Gur, Ruben C; Verma, Ragini

    2011-09-15

    Facial expression is widely used to evaluate emotional impairment in neuropsychiatric disorders. Ekman and Friesen's Facial Action Coding System (FACS) encodes movements of individual facial muscles from distinct momentary changes in facial appearance. Unlike facial expression ratings based on categorization of expressions into prototypical emotions (happiness, sadness, anger, fear, disgust, etc.), FACS can encode ambiguous and subtle expressions, and therefore is potentially more suitable for analyzing the small differences in facial affect. However, FACS rating requires extensive training, and is time consuming and subjective thus prone to bias. To overcome these limitations, we developed an automated FACS based on advanced computer science technology. The system automatically tracks faces in a video, extracts geometric and texture features, and produces temporal profiles of each facial muscle movement. These profiles are quantified to compute frequencies of single and combined Action Units (AUs) in videos, and they can facilitate a statistical study of large populations in disorders known to impact facial expression. We derived quantitative measures of flat and inappropriate facial affect automatically from temporal AU profiles. Applicability of the automated FACS was illustrated in a pilot study, by applying it to data of videos from eight schizophrenia patients and controls. We created temporal AU profiles that provided rich information on the dynamics of facial muscle movements for each subject. The quantitative measures of flatness and inappropriateness showed clear differences between patients and the controls, highlighting their potential in automatic and objective quantification of symptom severity.

  19. Performance analysis of electronic structure codes on HPC systems: A case study of SIESTA

    CERN Document Server

    Corsetti, Fabiano

    2014-01-01

    We report on scaling and timing tests of the SIESTA electronic structure code for ab initio molecular dynamics simulations using density-functional theory. The tests are performed on six large-scale supercomputers belonging to the PRACE Tier-0 network with four different architectures: Cray XE6, IBM BlueGene/Q, BullX, and IBM iDataPlex. We employ a systematic strategy for simultaneously testing hard and soft scaling, and propose a measure which is independent of the range of number of cores on which the tests are performed to quantify hard scaling efficiency as a function of simulation size. We find an increase in efficiency with simulation size for all machines, with a qualitatively different curve depending on the supercomputer topology, and discuss the connection of this functional form with soft scaling behaviour. We also analyze the absolute timings obtained in our tests, showing the range of system sizes and cores favourable for different machines. Our results can be employed as a guide both for running...

  20. HAMMER code system

    International Nuclear Information System (INIS)

    The development of a high-accuracy reactor benchmark analysis capability is described. This capability has been incorporated into a revised and extended version of the lattice analysis program HAMMER. Previous analyses using the HAMMER program required the introduction of correction factors obtained from more rigorous treatments of various effects such as resonance capture and neutron leakage. The present version of the program will remove the ambiguities associated with the introduction of such correction factors by optionally performing the more rigorous calculations internally or by automating the correctional procedure

  1. Speckle revisited: analysis of speckle noise in bar-code scanning systems

    Science.gov (United States)

    Marom, Emanuel; Kresic-Juric, Sasa; Bergstein, Leonard

    2001-06-01

    Laser beams used for bar-code scanning exhibit speckle noise generated by the roughness of the surface on which bar-codes are printed. Statistical properties of a photodetector signal that integrates a time-varying speckle pattern falling on its aperture are analyzed in detail. We derive simple closed form expressions for the auto-correlation function and power spectral density of the detector current for general form scanning beams with arbitrary field distributions. Theoretical calculations are illustrated by numerical simulations.

  2. Stability analysis by ERATO code

    International Nuclear Information System (INIS)

    Problems in MHD stability calculations by ERATO code are described; which concern convergence property of results, equilibrium codes, and machine optimization of ERATO code. It is concluded that irregularity on a convergence curve is not due to a fault of the ERATO code itself but due to inappropriate choice of the equilibrium calculation meshes. Also described are a code to calculate an equilibrium as a quasi-inverse problem and a code to calculate an equilibrium as a result of a transport process. Optimization of the code with respect to I/O operations reduced both CPU time and I/O time considerably. With the FACOM230-75 APU/CPU multiprocessor system, the performance is about 6 times as high as with the FACOM230-75 CPU, showing the effectiveness of a vector processing computer for the kind of MHD computations. This report is a summary of the material presented at the ERATO workshop 1979(ORNL), supplemented with some details. (author)

  3. ELCOS: the PSI code system for LWR core analysis. Part II: user`s manual for the fuel assembly code BOXER

    Energy Technology Data Exchange (ETDEWEB)

    Paratte, J.M.; Grimm, P.; Hollard, J.M. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-02-01

    ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user`s manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs.

  4. Error Correction Coding Meets Cyber-Physical Systems: Message-Passing Analysis of Self-Healing Interdependent Networks

    CERN Document Server

    Behfarnia, Ali

    2016-01-01

    Coupling cyber and physical systems gives rise to numerous engineering challenges and opportunities. An important challenge is the contagion of failure from one system to another, which can lead to large-scale cascading failures. However, the \\textit{self-healing} ability emerges as a valuable opportunity where the overlaying cyber network can cure failures in the underlying physical network. To capture both self-healing and contagion, this paper considers a graphical model representation of an interdependent cyber-physical system, in which nodes represent various cyber or physical functionalities, and edges capture the interactions between the nodes. A message-passing algorithm used in low-density parity-check codes is extended to this representation to study the dynamics of failure propagation and healing. By applying a density evolution analysis to this algorithm, network reaction to initial disruptions is investigated. It is proved that as the number of message-passing iterations increases, the network re...

  5. Intermittency coding in the primary olfactory system: a neural substrate for olfactory scene analysis.

    Science.gov (United States)

    Park, Il Memming; Bobkov, Yuriy V; Ache, Barry W; Príncipe, José C

    2014-01-15

    The spatial and temporal characteristics of the visual and acoustic sensory input are indispensable attributes for animals to perform scene analysis. In contrast, research in olfaction has focused almost exclusively on how the nervous system analyzes the quality and quantity of the sensory signal and largely ignored the spatiotemporal dimension especially in longer time scales. Yet, detailed analyses of the turbulent, intermittent structure of water- and air-borne odor plumes strongly suggest that spatio-temporal information in longer time scales can provide major cues for olfactory scene analysis for animals. We show that a bursting subset of primary olfactory receptor neurons (bORNs) in lobster has the unexpected capacity to encode the temporal properties of intermittent odor signals. Each bORN is tuned to a specific range of stimulus intervals, and collectively bORNs can instantaneously encode a wide spectrum of intermittencies. Our theory argues for the existence of a novel peripheral mechanism for encoding the temporal pattern of odor that potentially serves as a neural substrate for olfactory scene analysis.

  6. An Analysis of Syndrome Coding

    Science.gov (United States)

    Amiruzzaman, Md; Abdullah-Al-Wadud, M.; Chung, Yoojin

    In this paper a detail analysis is presented based on BCH syndrome coding for covert channel data hiding methods. The experimented technique is nothing but a syndrome coding algorithm with a coset based approach, analyzed results are showing that the examined method has more flexibility to choose coset, also providing less modification distortion caused by data hiding. Analyzed method presented by clear mathematical way. As it is mathematical equation dependent, hence analyzed results are showing that the analyzed method has fast computation ability and find perfect roots for modification.

  7. Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems, including containment. Summary report of a technical meeting

    International Nuclear Information System (INIS)

    Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The publication constitutes the report of the Technical Meeting. It includes short summaries of the presentations that were made and of the discussions as well as conclusions and

  8. Theoretical analysis and simulation of a code division multiple access system (cdma for secure signal transmission in wideband channels

    Directory of Open Access Journals (Sweden)

    Stevan M. Berber

    2014-06-01

    Code Division Multiple Access (CDMA technique which allows communications of multiple users in the same communication system. This is achieved in such a way that each user is assigned a unique code sequence, which is used at the receiver side to discover the information dedicated to that user. These systems belong to the group of communication systems for direct sequence spread spectrum systems. Traditionally, CDMA systems use binary orthogonal spreading codes. In this paper, a mathematical model and simulation of a CDMA system based on the application of non-binary, precisely speaking, chaotic spreading sequences. In their nature, these sequences belong to random sequences with infinite periodicity, and due to that they are appropriate for applications in the systems that provide enhanced security against interception and secrecy in signal transmission. Numerous papers are dedicated to the development of CDMA systems in flat fading channels. This paper presents the results of these systems analysis for the case when frequency selective fading is present in the channel. In addition, the paper investigates a possibility of using interleaving techniques to mitigate fading in a wideband channel with the frequency selective fading. Basic structure of a CDMA communication system and its operation In this paper, a CDMA system block schematic is ppresented and the function of all blocks is explained. Notation  to be used in the paper is introduced. Chaotic sequences are defined and explained in accordance with the method of their generation. A wideband channel with frequency selective fading is defined by its impulse response function. Theoretical analysis of a CDMA system with flat fading in a narrowband channel A narrowband channel and flat fading are defined. A mathematical analysis of the system is conducted by presenting the signal expressions at vital points in the transmitter and receiver. The expression of the signal at the output of the sequence correlator is

  9. Thermal-hydraulic system analysis using the MARS code for the transient steam generator tube rupture accident

    International Nuclear Information System (INIS)

    A postulated SGTR accident of the APR1400 was analysed using the best estimate safety analysis code, MARS. The main objective of this study is not only to provide physical insight into the system response of the APR1400 reactor during a SGTR but also to investigate the effect of reactor trip type of a HSGL and a LPP on the thermal-hydraulic system response. As for the tube rupture modelling method, double tube modelling was adopted. Broken U-tubes were modelled as a separate assembly of a single volume. The reactor trip type affected the overall progress of the major events. However, the effect on the thermal-hydraulic response of the plant was trivial. (author)

  10. Description of the TREBIL, CRESSEX and STREUSL computer programs, that belongs to RALLY computer code pack for the analysis of reliability systems

    International Nuclear Information System (INIS)

    The RALLY computer code pack (RALLY pack) is a set of computer codes destinate to the reliability of complex systems, aiming to a risk analysis. Three of the six codes, are commented, presenting their purpose, input description, calculation methods and results obtained with each one of those computer codes. The computer codes are: TREBIL, to obtain the fault tree logical equivalent; CRESSEX, to obtain the minimal cut and the punctual values of the non-reliability and non-availability of the system; and STREUSL, for the dispersion calculation of those values around the media. In spite of the CRESSEX, in its version available at CNEN, uses a little long method to obtain the minimal cut in an HB-CNEN system, the three computer programs show good results, mainly the STREUSL, which permits the simulation of various components. (E.G.)

  11. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Code Reference Manual

    Energy Technology Data Exchange (ETDEWEB)

    C. L. Smith; K. J. Kvarfordt; S. T. Wood

    2006-07-01

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer. However, the INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users comprised of a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for ansforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with

  12. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Code Reference Manual

    Energy Technology Data Exchange (ETDEWEB)

    C. L. Smith; K. J. Kvarfordt; S. T. Wood

    2008-08-01

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer. However, the INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users comprised of a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for transforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with

  13. Numerical Zooming Between a NPSS Engine System Simulation and a One-Dimensional High Compressor Analysis Code

    Science.gov (United States)

    Follen, Gregory; auBuchon, M.

    2000-01-01

    Within NASA's High Performance Computing and Communication (HPCC) program, NASA Glenn Research Center is developing an environment for the analysis/design of aircraft engines called the Numerical Propulsion System Simulation (NPSS). NPSS focuses on the integration of multiple disciplines such as aerodynamics, structures, and heat transfer along with the concept of numerical zooming between zero-dimensional to one-, two-, and three-dimensional component engine codes. In addition, the NPSS is refining the computing and communication technologies necessary to capture complex physical processes in a timely and cost-effective manner. The vision for NPSS is to create a "numerical test cell" enabling full engine simulations overnight on cost-effective computing platforms. Of the different technology areas that contribute to the development of the NPSS Environment, the subject of this paper is a discussion on numerical zooming between a NPSS engine simulation and higher fidelity representations of the engine components (fan, compressor, burner, turbines, etc.). What follows is a description of successfully zooming one-dimensional (row-by-row) high-pressure compressor analysis results back to a zero-dimensional NPSS engine simulation and a discussion of the results illustrated using an advanced data visualization tool. This type of high fidelity system-level analysis, made possible by the zooming capability of the NPSS, will greatly improve the capability of the engine system simulation and increase the level of virtual test conducted prior to committing the design to hardware.

  14. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author)

  15. Performance Analysis of Video Frame Transmission on DVB-H 4K Mode System for different code rates

    Directory of Open Access Journals (Sweden)

    Mitul Prajapati

    2012-03-01

    Full Text Available DVB-H (Digital Video Broadcasting for Handheld terminals is a digital transmission standard developed by the international DVB-Project. It was standardized in 2004 and enables small battery powered handheld devices to receive IP data services such as low definition TV services. The DVB-H standard is derived from the DVB-T standard which is used to broadcast TV services in Europe. The convolution code is used as error correction code. In this paper, we have implemented physical layer of DVB-H for 4K mode system. Here we had transmitted a JPEG format of frame of movie through AWGN channel and observed the image quality for different code rates used for convolution coding. The code rates used are 1/2, 2/3, 3/4, 5/6 and 7/8.

  16. Conceptual design study of a superconducting spherical tokamak reactor with a self-consistent system analysis code

    International Nuclear Information System (INIS)

    In a spherical tokamak (ST) reactor, the radial build of toroidal field coil and the shield play a key role in determining the size of the reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with a one-dimensional radiation transport code. A conceptual design study of a compact superconducting ST reactor with an aspect ratio of up to 2.0 is conducted and the optimum radial build is identified. It is shown that the use of an improved shielding material and high-temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at a low aspect ratio, and that by using an inboard neutron reflector instead of a breeding blanket, tritium self-sufficiency is possible with an outboard blanket only and thus a compact-sized all superconducting coil ST reactor is viable.

  17. The Performance Analysis of Traffic Channel Coding in Digital Trunking System

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The encoding and decoding processes of traffic channel in digital trunking system are studied. On the basis of computer simulation, the BER (bit error ratio) with different RCPC decoding step is analyzed. As a result, the optimal RCPC decoding step is provided, which gives essential theoretical evidences for the implementation of digital trunking system.

  18. Analysis of Hydrogen Risk Mitigation System for Severe Accidents of EU-APR1400 Using MAAP4 code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Mun Soo; Suh, Jung Soo; Bae, Byoung Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    According to the EUR (European Utility Requirements for LWR Nuclear Power Plants), it is mandatory that the HMS (Hydrogen Mitigation System) of the Eu-APR1400 should be equipped with a passive or automatic hydrogen control system. Considering this requirement, a PAR (Passive Autocatalytic Recombiner) system was adopted for the HMS of the Eu-APR1400. This passive HMS should be evaluated carefully in order to ensure that the HMS has adequate capacity to control hydrogen concentrations during severe accident conditions and to show that the system can satisfy the design requirements of the EUR. In this paper, analyses were carried out to examine the effectiveness of the HMS incorporated into the Eu- APR1400 design. These analyses were performed using the MAAP (Modular Accident Analysis Program) 4 code. in order to identify whether the HMS could control the average hydrogen concentrations in the containment, such that the concentration would not exceed 10 percent by volume: the analyses also considered whether there was the possibility of inadvertent hydrogen combustion in such processes as FA (Flame Acceleration) and DDT (Deflagration to Detonation Transition)

  19. The solution of linear systems of equations with a structural analysis code on the NAS CRAY-2

    Science.gov (United States)

    Poole, Eugene L.; Overman, Andrea L.

    1988-01-01

    Two methods for solving linear systems of equations on the NAS Cray-2 are described. One is a direct method; the other is an iterative method. Both methods exploit the architecture of the Cray-2, particularly the vectorization, and are aimed at structural analysis applications. To demonstrate and evaluate the methods, they were installed in a finite element structural analysis code denoted the Computational Structural Mechanics (CSM) Testbed. A description of the techniques used to integrate the two solvers into the Testbed is given. Storage schemes, memory requirements, operation counts, and reformatting procedures are discussed. Finally, results from the new methods are compared with results from the initial Testbed sparse Choleski equation solver for three structural analysis problems. The new direct solvers described achieve the highest computational rates of the methods compared. The new iterative methods are not able to achieve as high computation rates as the vectorized direct solvers but are best for well conditioned problems which require fewer iterations to converge to the solution.

  20. Computer access security code system

    Science.gov (United States)

    Collins, Earl R., Jr. (Inventor)

    1990-01-01

    A security code system for controlling access to computer and computer-controlled entry situations comprises a plurality of subsets of alpha-numeric characters disposed in random order in matrices of at least two dimensions forming theoretical rectangles, cubes, etc., such that when access is desired, at least one pair of previously unused character subsets not found in the same row or column of the matrix is chosen at random and transmitted by the computer. The proper response to gain access is transmittal of subsets which complete the rectangle, and/or a parallelepiped whose opposite corners were defined by first groups of code. Once used, subsets are not used again to absolutely defeat unauthorized access by eavesdropping, and the like.

  1. Understanding Code Patterns - Analysis, Interpretation & Measurement

    CERN Document Server

    Dundas, Jitesh

    2011-01-01

    This research paper aims to find, analyze and understand code patterns in any software system and measure its quality by defining standards and proposing a formula for the same. Every code that is written can be divided into different code segments, each having its own impact on the overall system. We can analyze these code segments to get the code quality. The measures used in this paper include Lines of Code, Number of calls made by a module, Execution time, the system knowledge of user and developers, the use of generalization, inheritance, reusability and other object-oriented concepts. The entire software code is divided into code snippets, based on the logic that they implement. Each of these code snippets has an impact. This measure is called Impact Factor and is valued by the software developer and/or other system stakeholders. Efficiency = (Code Area / Execution Time) * Qr

  2. The stability analysis using two fluids (SAT/trademark/) code for boiling flow systems: Volume 3: User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.P.; Dykhuizen, R.C.; Su, M.G.; Jain, P.

    1988-12-01

    This report presents analyses of dynamic instability and frequency response characteristics of boiling flow systems based on an unequal velocity, unequal temperature two-fluid model of such flow. The dynamic instability analyses in the time domain are incorporated into three options of a computer code SAT, viz., DI01 (steady state, or equilibrium point analysis), DI02 (linear stability analysis in time domain), and DI03 (nonlinear analysis in time domain). The frequency response analysis is incorporated into a fourth option FREQ. Results of dynamic instability experiments carried out in a Refrigerant-113 boiling flow rig are also reported as are comparison of these with linear stability analysis predictions. A description of the input file structure of the four codes is present in this volume of the report. Outputs of these codes are also described in detail. Sample input and output files are included in the appendices of this volume.

  3. Development of Pool-type Sodium-cooled Fast Reactor System Analysis Code%池式钠冷快堆系统分析程序开发

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

    针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用 FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序(FASYS程序)。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。%According to the characteristics of pool‐type sodium‐cooled fast reactor ,and with the fast reactor hydraulic model , thermal model and neutron kinetics model thoroughly classified and developed ,a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool‐type sodium‐cooled fast reactor acci‐dent analysis .Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation .The calculation results are consistent with the test data and other fast reactor system analysis code results , and the correctness of the FASYS code is proved .

  4. Sandia National Laboratories analysis code data base

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, C.W.

    1994-11-01

    Sandia National Laboratories, mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The Laboratories` strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia`s technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code ``ownership`` and release status, and references describing the physical models and numerical implementation.

  5. Web interface for plasma analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Emoto, M. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)], E-mail: emo@nifs.ac.jp; Murakami, S. [Kyoto University, Yoshida-Honmachi, Sakyo-ku, Kyoto 606-8501 (Japan); Yoshida, M.; Funaba, H.; Nagayama, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)

    2008-04-15

    There are many analysis codes that analyze various aspects of plasma physics. However, most of them are FORTRAN programs that are written to be run in supercomputers. On the other hand, many scientists use GUI (graphical user interface)-based operating systems. For those who are not familiar with supercomputers, it is a difficult task to run analysis codes in supercomputers, and they often hesitate to use these programs to substantiate their ideas. Furthermore, these analysis codes are written for personal use, and the programmers do not expect these programs to be run by other users. In order to make these programs to be widely used by many users, the authors developed user-friendly interfaces using a Web interface. Since the Web browser is one of the most common applications, it is useful for both the users and developers. In order to realize interactive Web interface, AJAX technique is widely used, and the authors also adopted AJAX. To build such an AJAX based Web system, Ruby on Rails plays an important role in this system. Since this application framework, which is written in Ruby, abstracts the Web interfaces necessary to implement AJAX and database functions, it enables the programmers to efficiently develop the Web-based application. In this paper, the authors will introduce the system and demonstrate the usefulness of this approach.

  6. Understanding Code Patterns - Analysis, Interpretation & Measurement

    OpenAIRE

    Dundas, Jitesh

    2011-01-01

    This research paper aims to find, analyze and understand code patterns in any software system and measure its quality by defining standards and proposing a formula for the same. Every code that is written can be divided into different code segments, each having its own impact on the overall system. We can analyze these code segments to get the code quality. The measures used in this paper include Lines of Code, Number of calls made by a module, Execution time, the system knowledge of user and...

  7. Thermal-hydraulic analysis for the lead-bismuth eutectic cooled reactor. System analysis by MSG-COPD code

    International Nuclear Information System (INIS)

    The feasibility study for fast breeder reactors (FBRs) including related nuclear fuel cycle systems has been started from the 1999 fiscal year by Japan Nuclear Cycle Development Institute (JNC). Phase 1 studies were finished at the end of March, 2000. Various options of FBRs plant systems was studied and concept of Lead-Bismuth Eutectic (LBE) cooled FBRs have been selected as one of these options. In the United States, the LBE cooled reactor has been examined by Generation IV. Plant dynamics analyses on 2 type of LBE-cooled reactors, forced circulation type which designed by JNC and natural circulation type which was designed by University of California, Berkeley, have been performed to understand the basic thermal-hydraulic characteristics of the reactors. As a result of the analysis on JNC forced circulation reactor, it has been clarified that hot coolant remains in the upper plenum by the thermal stratification in case of a manual trip condition. And the characteristics of pump coast down influences core exit high-temperature in case of a loss of power condition. In addition, as a result of analysis on the natural circulation reactor, the flow-redistribution effect in ductless core channels by the buoyancy force has been evaluated for a candidate duct core channels. (author)

  8. Expansion of the CHR bone code system

    International Nuclear Information System (INIS)

    This report describes the coding system used in the Center for Human Radiobiology (CHR) to identify individual bones and portions of bones of a complete skeletal system. It includes illustrations of various bones and bone segments with their respective code numbers. Codes are also presented for bone groups and for nonbone materials

  9. Performance Analysis of Global Search Algorithm Based Multiuser Detector for Multi Carrier Code Division Multiple Access System under Clipping Noise

    Directory of Open Access Journals (Sweden)

    S. Sivanesskumar

    2011-01-01

    Full Text Available Problem statement: Multi Carrier Code Division Multiple Access (MC-CDMA, a promising technology for the 4G communication systems. The major limitation of MC-CDMA system is the Multiple Access Interference (MAI which is due to near-far effect, frequency offset and nonlinear power amplification due to clipping noise. Approach: The performance of MC-CDMA under clipping noise using Global search algorithm based Multiuser detector in AWGN, Rayleigh and Rician channel is analyzed in this study. Results: The proposed method is simulated using BPSK modulation, Walsh spreading code, number of subcarriers 64 and number of users 16 and clipping noise. Conclusion: By simulation result, BER in AWGN channel outperforms other channels as SNR is increased. The performance of Rician fading channel is better than that of Rayleigh fading channel, because of the LOS path.

  10. Recent developments in the Los Alamos radiation transport code system

    Energy Technology Data Exchange (ETDEWEB)

    Forster, R.A.; Parsons, K. [Los Alamos National Lab., NM (United States)

    1997-06-01

    A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results.

  11. Characterizing Video Coding Computing in Conference Systems

    NARCIS (Netherlands)

    Tuquerres, G.

    2000-01-01

    In this paper, a number of coding operations is provided for computing continuous data streams, in particular, video streams. A coding capability of the operations is expressed by a pyramidal structure in which coding processes and requirements of a distributed information system are represented. Th

  12. Development of breached pin performance analysis code SAFFRON (System of Analyzing Failed Fuel under Reactor Operation by Numerical method)

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

    On the assumption of fuel pin failure, the breached pin performance analysis code SAFFRON was developed to evaluate the fuel pin behavior in relation to the delayed neutron signal response during operational mode beyond the cladding failure. Following characteristic behavior in breached fuel pin is modeled in 3-dimensional finite element method : pellet swelling by fuel-sodium reaction, fuel temperature change, and resultant cladding breach extension and delayed neutron precursors release into coolant. Particularly, practical algorithm of numerical procedure in finite element method was originally developed in order to solve the 3-dimensional non-linear contact problem between the swollen pellet due to fuel-sodium reaction and breached cladding. (author).

  13. Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

    International Nuclear Information System (INIS)

    The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain

  14. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)

  15. Wind energy systems application to regional utilities. [SERIES code; WINDS code; PHASES code; AVERAGE code; NETLOAD code; GENSYS code; PROCOST code; CAP6 code; EVEN code

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    A methodology for analyzing the economic impact of WECS on a utility is described in Volume I of this report. The methodology requires extrapolating both historical utility load data and historical wind power into a year of analysis; calculating the total amount of funds made available in that year, as a result of the inclusion of wind power in the utility mix; and then estimating the present value of the total funds made available to the utility over the life of the WECS. To apply the methodology to a specific case, it was necessary to develop various computer programs. The following sections in this report list the programs developed for this study, briefly summarize their contents, and explain how they are used. Wherever possible, a typical input/output file is shown.

  16. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

  17. RDS; A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    International Nuclear Information System (INIS)

    Full-text: Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBIMOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges. (author)

  18. MARS code manual volume I: code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  19. Transmission Analysis of Optical Code Division Multiple Access Communication Systems in the Presence of Noise in Local Area Network Applications

    Directory of Open Access Journals (Sweden)

    Ahmed Nabih Zaki Rashed

    2013-04-01

    Full Text Available OCDMA is an essential part of the digital communication system now days for long haul, high speed networks. The biggest challenge with Optical CDMA system is to maintain the performance of the system and offer high bandwidth in case of higher number of users at minimum cost. As the number of users increase, the input requirements i.e. transmitted power, bit rate etc start increasing sharply which contribute to the additional cost. It has recently attracted significant research interest because of the advantages it offers in terms of the flexibility in the management of the system resources. We have taken into account the system design parameters are determined such as BER (bit error rate, signal to noise ratio (SNR, transmission bit rates, and optical received power for different code lengths. The Optical CDMA systems suffer from the problem of multiple access interference (MAI.As the number of users increase the BER error rate degrades because the effect of MAI (multiple access interference increases. So, there is a limitation in number of users, as the number of users increase SNR decrease and probability of error increases.

  20. The EGS5 Code System

    Energy Technology Data Exchange (ETDEWEB)

    Hirayama, Hideo; Namito, Yoshihito; /KEK, Tsukuba; Bielajew, Alex F.; Wilderman, Scott J.; U., Michigan; Nelson, Walter R.; /SLAC

    2005-12-20

    In the nineteen years since EGS4 was released, it has been used in a wide variety of applications, particularly in medical physics, radiation measurement studies, and industrial development. Every new user and every new application bring new challenges for Monte Carlo code designers, and code refinements and bug fixes eventually result in a code that becomes difficult to maintain. Several of the code modifications represented significant advances in electron and photon transport physics, and required a more substantial invocation than code patching. Moreover, the arcane MORTRAN3[48] computer language of EGS4, was highest on the complaint list of the users of EGS4. The size of the EGS4 user base is difficult to measure, as there never existed a formal user registration process. However, some idea of the numbers may be gleaned from the number of EGS4 manuals that were produced and distributed at SLAC: almost three thousand. Consequently, the EGS5 project was undertaken. It was decided to employ the FORTRAN 77 compiler, yet include as much as possible, the structural beauty and power of MORTRAN3. This report consists of four chapters and several appendices. Chapter 1 is an introduction to EGS5 and to this report in general. We suggest that you read it. Chapter 2 is a major update of similar chapters in the old EGS4 report[126] (SLAC-265) and the old EGS3 report[61] (SLAC-210), in which all the details of the old physics (i.e., models which were carried over from EGS4) and the new physics are gathered together. The descriptions of the new physics are extensive, and not for the faint of heart. Detailed knowledge of the contents of Chapter 2 is not essential in order to use EGS, but sophisticated users should be aware of its contents. In particular, details of the restrictions on the range of applicability of EGS are dispersed throughout the chapter. First-time users of EGS should skip Chapter 2 and come back to it later if necessary. With the release of the EGS4 version

  1. Spectral Analysis Code: PARAS SPEC

    CERN Document Server

    Chaturvedi, Priyanka; Anandarao, B G

    2016-01-01

    The light emitted from the stellar photosphere serves as a unique signature for the nature of stars. The behaviour of these stellar lines depend upon the surface temperature, mass, evolutionary status and chemical composition of the star. With the advent of high-resolution spectrographs coupled with medium to large aperture telescopes around the globe, there is plenty of high-resolution and high signal-to-noise ratio data available to the astronomy community. Apart from radial velocity (RV) studies, such data offer us the unique opportunity to study chemical composition and atmospheric properties of the star. The procedure used to derive these parameters must be automated and well adaptable to data available from any high-resolution spectrograph. We hereby present an IDL code, PARAS SPEC, which was primary designed to handle high-resolution spectroscopy data from PARAS spectrograph coupled with the 1.2~m telescope at Mt. Abu, India. This code is designed to adapt with data from other spectrographs as well. Th...

  2. SCALE Code System 6.2.1

    Energy Technology Data Exchange (ETDEWEB)

    Rearden, Bradley T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jessee, Matthew Anderson [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

  3. User's manual for seismic analysis code 'SONATINA-2V'

    International Nuclear Information System (INIS)

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

  4. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

    Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods

  5. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  6. Software and codes for analysis of concentrating solar power technologies.

    Energy Technology Data Exchange (ETDEWEB)

    Ho, Clifford Kuofei

    2008-12-01

    This report presents a review and evaluation of software and codes that have been used to support Sandia National Laboratories concentrating solar power (CSP) program. Additional software packages developed by other institutions and companies that can potentially improve Sandia's analysis capabilities in the CSP program are also evaluated. The software and codes are grouped according to specific CSP technologies: power tower systems, linear concentrator systems, and dish/engine systems. A description of each code is presented with regard to each specific CSP technology, along with details regarding availability, maintenance, and references. A summary of all the codes is then presented with recommendations regarding the use and retention of the codes. A description of probabilistic methods for uncertainty and sensitivity analyses of concentrating solar power technologies is also provided.

  7. Code Formal Verification of Operation System

    Directory of Open Access Journals (Sweden)

    Yu Zhang

    2010-12-01

    Full Text Available with the increasing pressure on non-function attributes (security, safety and reliability requirements of an operation system, high–confidence operation system is becoming more important. Formal verification is the only known way to guarantee that a system is free of programming errors. We research on formal verification of operation system kernel in system code level and take theorem proving and model checking as the main technical methods to resolve the key techniques of verifying operation system kernel in C code level. We present a case study to the verification of real-world C systems code derived from an implementation of μC/OS – II in the end.

  8. Design of a physical format coding system

    Science.gov (United States)

    Hu, Beibei; Pei, Jing; Zhang, Qicheng; Liu, Hailong; Tang, Yi

    2008-12-01

    A novel design of physical format coding system (PFCS) is presented based on Multi-level read-only memory disc (ML ROM) in order to solve the problem of low efficiency and long period of disc testing during system development. The PFCS is composed of four units, which are 'Encode', 'Add Noise', 'Decode', 'Error Rate', and 'Information'. It is developed with MFC under the environment of VC++ 6.0, and capable to visually simulate the procedure of data processing for ML ROM. This system can also be used for developing other optical disc storage system or similar channel coding system.

  9. Uncertainty analysis of the SWEPP PAN assay system for glass waste (content codes 440, 441 and 442)

    Energy Technology Data Exchange (ETDEWEB)

    Blackwood, L.G.; Harker, Y.D.; Meachum, T.R.; Yoon, W.Y.

    1996-10-01

    INEL is being used as a temporary storage facility for transuranic waste generated by the Nuclear Weapons program at the Rocky Flats Plant. Currently, there is a large effort in progress to prepare to ship this waste to WIPP. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Action Neutron (PAN) radioassay system. This paper discusses a modified statistical sampling and verification approach used to determine the total uncertainty of SWEPP PAN measurements for glass waste (content codes 440, 441, and 442) contained in 208 liter drums. In the modified statistical sampling and verification approach, the total performance of the SWEPP PAN nondestructive assay system for specifically selected waste conditions is simulated using computer models. A set of 100 cases covering the known conditions exhibited in glass waste was compiled using a combined statistical sampling and factorial experimental design approach. Parameter values assigned in each simulation were derived from reviews of approximately 100 real-time radiography video tapes of RFP glass waste drums, results from previous SWEPP PAN measurements on glass waste drums, and shipping data from RFP where the glass waste was generated. The data in the 100 selected cases form the multi-parameter input to the simulation model. The reported plutonium masses from the simulation model are compared with corresponding input masses. From these comparisons, the bias and total uncertainty associated with SWEPP PAN measurements on glass waste drums are estimated. The validity of the simulation approach is verified by comparing simulated output against results from calibration measurements using known plutonium sources and two glass waste calibration drums.

  10. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)

  11. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  12. Modernization and restructuring of realistic thermal hydraulic system analysis code, RELAP5/MOD3.3.1.2

    International Nuclear Information System (INIS)

    The code architecture entails the programming language and the code database. Various recent programming languages such as C, C++, Fortran 90, were considered as the candidate language for the modernization of RELAP5/MOD3.2.1.2. Among them, Fortran 90 was selected as a basic programming laguage for the modernization and restructuring of the code. Most of header file (*.h) and equivalenced variables in RELAP5 have been replaced with members in the MODULE, which greatly enhance the code maintenance and readability. The FTB package is used for the dynamic memory management (DMM) of RELAP5. Although FTB DMM features are very successful, the use of FTB has been the obstacle in the maintenance of the code. It is difficult to understand and change the coding, and it requires a significant effort to find out index errors in large memory pools. With new features introduced in Fortran 90, it is possible to slove dynamic allocation problems within the standard features in an elegant, clear safe way. Each of FTB data blocks can be replaced by the suitably organized derived variables in MODULE and the standard DMM scheme. This DMM scheme provides the code flexibility which can save the memory requirements depending on the problem sizes without a extensive use of the complex FTB package. The current user's interface of the RELAP5 consists of a set of input file, output file, and restart/plot file. Many users complain that this interface is not user friendly. It was mainly caused by the text-oriented programming, namly console programming during the past many years. Now, windows programming has become popular in most areas of software development. Using this windows programming technique, the user friend freatures can be implemented. The Visual Fortran Quick Win run-time library helps to turn graphics programs into simple Windows applications. RELAP5 code has been re-compiled with the Quick Win feature, and the mask for user's dialog and graphical x-y plot were designed. This

  13. Ocean Thermal Energy Conversion power system development. Phase I: preliminary design. Final report. [OSAP-1 code; OTEC Steady-State Analysis Program

    Energy Technology Data Exchange (ETDEWEB)

    Westerberg, Arthur

    1978-12-04

    The following appendices are included: highlights of direction and correspondence; user manual for OTEC Steady-State Analysis Program (OSAP-1); sample results of OSAP-1; surface condenser installations; double-clad systems; aluminum alloy seawater piping; references searched for ammonia evaluation; references on stress-corrosion for ammonia; references on anhydrous ammonia storage; references on miscellaneous ammonia items; OTEC materials testing; test reports; OTEC technical specification chlorination system; OTEC technical specification AMERTAP system; OTEC optimization program users guide; concrete hull construction; weight and stability estimates; packing factor data; machinery and equipment list; letter from HPTI on titanium tubes; tables on Wolverine Korodense tubes; evaporator and condenser enhancement tables; code weld titanium tube price, weight tables Alcoa tubing tables; Union Carbide tubing pricing tables; turbotec tubing pricing tables; Wolverine tubing pricing tables; Union Carbide tubing characteristics and pricing; working fluids and turbines for OTEC power system; and hydrodynamic design of prototype OTEC cold and warm seawater pumps. (WHK)

  14. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  15. Analysis of MSGTR events for APR1400 by means of best estimate thermal-hydraulic system code

    International Nuclear Information System (INIS)

    A multiple steam generator tube rupture (MSGTR) event has never occurred in the history of commercial nuclear reactor operation while single steam generator tube rupture (SGTR) event is reported to occur every two years. As there is no history of MSGTR event, the understandings of transients and consequences of this event are not so much. In this study, a postulated MSGTR event in advanced power reactor 1400 (APR1400) is analyzed using thermal-hydraulic system code. The APR 1400 is a two-loop, 1000 MWe, PWR supposed to be built in 2009. MARS1.4 is used in this study. The present study aims to understand the effects of rupture location in heat transfer tubes and selection of affected steam generator following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 is to allow shortest time for operator action following a tubes rupture in the vicinity of hot-leg side tube sheet and to allow longest time following a tube ruptures at the tube top. The MSSV lift time for rupture at tube-top is evaluated as 24.5% larger than that for rupture at hot-leg side tube sheet. Also, the MSSV lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generator is affected. The comparison shows that the cases for both of two steam generators are affected allow longer time for operator action compared with the cases that a single steam generator is affected. Further more, the tube ruptures in the steam generator where a pressurizer is linked leads to the shortest operator response time

  16. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  17. Analysis of core physics test data and sodium void reactivity worth calculation for MONJU core with ARCADIAN-FBR computer code system

    International Nuclear Information System (INIS)

    In order to evaluate core characteristics of fast reactors, a computer code system ARCADIAN-FBR has been developed by utilizing the existing analysis codes and the latest nuclear data library JENDL-3.3. The validity of ARCADIAN-FBR was verified by using the experimental data obtained in the MONJU core physics tests. The results of analyses are in good agreement with the experimental data and the applicability of ARCADIAN-FBR for fast reactor core analysis is confirmed. Using ARCADIAN-FBR, the sodium void reactivity worth, which is an important parameter in the safety analysis of fast reactors, was analyzed for MONJU core. 241Pu in the core fuel is transmuted to 241Am due to disintegrations. Therefore, the effect of 241Am accumulation on the sodium void reactivity worth was evaluated for MONJU core. As a result of calculation, it was confirmed that the accumulation of 241Am significantly influences on the sodium void reactivity worth and hence on the safety analysis of sodium-cooled fast reactors. (author)

  18. Development and improvement of safety analysis code for geological disposal

    International Nuclear Information System (INIS)

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  19. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  20. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs

  1. Analysis of Void Fraction Distribution and Departure from Nucleate Boiling in Single Subchannel and Bundle Geometries Using Subchannel, System, and Computational Fluid Dynamics Codes

    OpenAIRE

    Taewan Kim; Victor Petrov; Annalisa Manera; Simon Lo

    2012-01-01

    In order to assess the accuracy and validity of subchannel, system, and computational fluid dynamics codes, the Paul Scherrer Institut has participated in the OECD/NRC PSBT benchmark with the thermal-hydraulic system code TRACE5.0 developed by US NRC, the subchannel code FLICA4 developed by CEA, and the computational fluid dynamic code STAR-CD developed by CD-adapco. The PSBT benchmark consists of a series of void distribution exercises and departure from nucleate boiling exercises. The resul...

  2. Code Formal Verification of Operation System

    OpenAIRE

    Yu Zhang; Yunwei Dong; Huo Hong; Fan Zhang

    2010-01-01

    with the increasing pressure on non-function attributes (security, safety and reliability) requirements of an operation system, high–confidence operation system is becoming more important. Formal verification is the only known way to guarantee that a system is free of programming errors. We research on formal verification of operation system kernel in system code level and take theorem proving and model checking as the main technical methods to resolve the key techniques of verifying operatio...

  3. Calculational results using a survey type code system for the analysis of the Almaraz Unit 2 PWR benchmark

    International Nuclear Information System (INIS)

    The calculations performed for the Almaraz Unit 2 PWR using the code packages of the Atomic Energy Corporation of South Africa Ltd. are summarized. These calculations were done as part of the IAEA Coordinated Research Programme on In-Core Fuel Management Code Package Validation for LWRs. A brief description of the one-dimensional cross section generation package as well as of the Level II (scoping type) global core calculational package which was used is given. Detailed results are presented in several appendices. 29 figs., 20 tabs., 10 refs

  4. CATHENA 4. A thermalhydraulics network analysis code

    International Nuclear Information System (INIS)

    Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA) is a one-dimensional, non-equilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. The objective of the present paper is to describe the design, application and future development plans for the CATHENA 4 thermalhydraulics network analysis code, which is a modernized version of the present frozen CATHENA 3 code. The new code is designed in modular form, using the Fortran 95 (F95) programming language. The semi-implicit numerical integration scheme of CATHENA 3 is re-written to implement a fully-implicit methodology using Newton's iterative solution scheme suitable for nonlinear equations. The closure relations, as a first step, have been converted from the existing CATHENA 3 implementation to F95 but modularized to achieve ease of maintenance. The paper presents the field equations, followed by a description of the Newton's scheme used. The finite-difference form of the field equations is given, followed by a discussion of convergence criteria. Two applications of CATHENA 4 are presented to demonstrate the temporal and spatial convergence of the new code for problems with known solutions or available experimental data. (author)

  5. Principles of the reactor code system RHEIN

    International Nuclear Information System (INIS)

    A description is given of the principles of the reactor code system RHEIN which is applied in connection with a BESM6-type computer. In transfering data between the components of the system external storage is used. The programme passage is controlled by the input data. (author)

  6. Static Code Analysis with Gitlab-CI

    CERN Document Server

    Datko, Szymon Tomasz

    2016-01-01

    Static Code Analysis is a simple but efficient way to ensure that application’s source code is free from known flaws and security vulnerabilities. Although such analysis tools are often coming with more advanced code editors, there are a lot of people who prefer less complicated environments. The easiest solution would involve education – where to get and how to use the aforementioned tools. However, counting on the manual usage of such tools still does not guarantee their actual usage. On the other hand, reducing the required effort, according to the idea “setup once, use anytime without sweat” seems like a more promising approach. In this paper, the approach to automate code scanning, within the existing CERN’s Gitlab installation, is described. For realization of that project, the Gitlab-CI service (the “CI” stands for "Continuous Integration"), with Docker assistance, was employed to provide a variety of static code analysers for different programming languages. This document covers the gene...

  7. MHD stability analysis code ERATO-J

    International Nuclear Information System (INIS)

    Necessary resources of a computer system for the MHD stability calculations by the ERATO are estimated. In this report, these data and concrete procedure to carry out a series of calculations by using the ERATO-J(F-version) code are described. The ERATO-H(F-version) is the first version of the ERATO code for the FACOM M200 computer system of JAERI computer center, which was adapted from the original ERATO code developed by R. Gruber et al. In this version several minor changes were introduced. Among them the DIARY program which facilitates acquisition and sorting of the output data is very useful to carry out a large amount of the ERATO calculations efficiently. (author)

  8. Finite-Length Analysis of BATS Codes

    OpenAIRE

    Yang, Shenghao; Ng, Tsz-Ching; Yeung, Raymond W.

    2013-01-01

    BATS codes were proposed for communication through networks with packet loss. A BATS code consists of an outer code and an inner code. The outer code is a matrix generation of a fountain code, which works with the inner code that comprises random linear coding at the intermediate network nodes. In this paper, the performance of finite-length BATS codes is analyzed with respect to both belief propagation (BP) decoding and inactivation decoding. Our results enable us to evaluate efficiently the...

  9. PROSA-1: a probabilistic response-surface analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vaurio, J. K.; Mueller, C.

    1978-06-01

    Techniques for probabilistic response-surface analysis have been developed to obtain the probability distributions of the consequences of postulated nuclear-reactor accidents. The uncertainties of the consequences are caused by the variability of the system and model input parameters used in the accident analysis. Probability distributions are assigned to the input parameters, and parameter values are systematically chosen from these distributions. These input parameters are then used in deterministic consequence analyses performed by mechanistic accident-analysis codes. The results of these deterministic consequence analyses are used to generate the coefficients for analytical functions that approximate the consequences in terms of the selected input parameters. These approximating functions are used to generate the probability distributions of the consequences with random sampling being used to obtain values for the accident parameters from their distributions. A computer code PROSA has been developed for implementing the probabilistic response-surface technique. Special features of the code generate or treat sensitivities, statistical moments of the input and output variables, regionwise response surfaces, correlated input parameters, and conditional distributions. The code can also be used for calculating important distributions of the input parameters. The use of the code is illustrated in conjunction with the fast-running accident-analysis code SACO to provide probability studies of LMFBR hypothetical core-disruptive accidents. However, the methods and the programming are general and not limited to such applications.

  10. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems

  11. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  12. SIMULATE-3K linkage with reactor systems codes

    International Nuclear Information System (INIS)

    SIMULATE-3K is Studsvik Scandpower's best-estimate three-dimensional core kinetics code. SIMULATE-3K has been coupled to several best-estimate reactor systems codes including, RELAP5-3D, RELAP5-3.3, TRACE V5.0, and RETRAN-3D. The coupled codes can be applied to existing reactors and to advanced reactor designs. The S3K linkage to each of the systems codes is a direct, explicit coupling of the two codes on a synchronous time-step basis. The coupling provides an execution method for the S3K three-dimensional neutronic model using the Nuclear Steam Supply System (NSSS) boundary conditions calculated by the systems code. Also, it allows the S3K calculated total core power and core power distributions to drive the system model core. Detailed calculations from the component codes result in a methodology for analyzing limiting transients such as steam line breaks, rod drops/ejections, and ATWS scenarios. These transient events require detailed three- dimensional core data and information about the behavior of NSSS components. A coupled analysis of these transients is important because the core behavior is closely tied to the NSSS system. For example, to capture the timing and characteristics of the important thermal-hydraulic phenomena and/or operations events, such as valve closures, safety injection, or control system interactions, requires a detailed plant model. The Peach Bottom 2 turbine trip transient is used to assess the accuracy of the coupled code calculations. Comparisons of the important plant parameters to results from RELAP5-3D, RELAP5-3.3, and TRACE V5.0 calculations are shown and discussed. The MSLB benchmark is also used to demonstrate the capabilities of the coupled code systems. Comparisons of the calculated reactor power to the reference data are shown can discussed. The comparisons demonstrate the applicability of S3K, either standalone or coupled with a system analysis code, to properly model system response during accident scenarios. (author)

  13. Innovations and enhancements in neutronic analysis of the Big-10 university research and training reactors based on the AGENT code system

    International Nuclear Information System (INIS)

    Introduction. This paper summarizes salient aspects of the 'virtual' reactor system developed at Purdue Univ. emphasizing efficient neutronic modeling through AGENT (Arbitrary Geometry Neutron Transport) a deterministic neutron transport code. DOE's Big-10 Innovations in Nuclear Infrastructure and Education (INIE) Consortium was launched in 2002 to enhance scholarship activities pertaining to university research and training reactors (URTRs). Existing and next generation URTRs are powerful campus tools for nuclear engineering as well as a number of disciplines that include, but are not limited to, medicine, biology, material science, and food science. Advancing new computational environments for the analysis and configuration of URTRs is an important Big-10 INIE aim. Specifically, Big-10 INIE has pursued development of a 'virtual' reactor, an advanced computational environment to serve as a platform on which to build operations, utilization (research and education), and systemic analysis of URTRs physics. The 'virtual' reactor computational system will integrate computational tools addressing the URTR core and near core physics (transport, dynamics, fuel management and fuel configuration); thermal-hydraulics; beam line, in-core and near-core experiments; instrumentation and controls; confinement/containment and security issues. Such integrated computational environment does not currently exist. The 'virtual' reactor is designed to allow researchers and educators to configure and analyze their systems to optimize experiments, fuel locations for flux shaping, as well as detector selection and configuration. (authors)

  14. Bar-code automated waste tracking system

    International Nuclear Information System (INIS)

    The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ''stop-and-go'' operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste

  15. Bar-code automated waste tracking system

    Energy Technology Data Exchange (ETDEWEB)

    Hull, T.E.

    1994-10-01

    The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ``stop-and-go`` operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste.

  16. Use of Serpent Monte-Carlo code for development of 3D full-core models of Gen-IV fast spectrum reactors and preparation of safety parameters/cross-section data for transient analysis with FAST code system

    International Nuclear Information System (INIS)

    Current work presents a new methodology which uses Serpent Monte-Carlo (MC) code for generating multi-group beginning-of-life (BOL) cross section (XS) database file that is compatible with PARCS 3D reactor core simulator and allows simulation of transients with the FAST code system. The applicability of the methodology was tested on European Sodium-cooled Fast Reactor (ESFR) design with an oxide fuel proposed by CEA (France). The k-effective, power peaking factors and safety parameters (such as Doppler constant, coolant density coefficient, fuel axial expansion coefficient, diagrid expansion coefficients and control rod worth) calculated by PARCS/TRACE were compared with the results of the Serpent MC code. The comparison indicates overall reasonable agreement between conceptually different (deterministic and stochastic) codes. The new development makes it in principle possible to use the Serpent MC code for cross section generation for the PARCS code to perform transient analyses for fast reactors. The advantages and limitations of this methodology are discussed in the paper. (author)

  17. User Instructions for the Systems Assessment Capability, Rev. 1, Computer Codes Volume 3: Utility Codes

    Energy Technology Data Exchange (ETDEWEB)

    Eslinger, Paul W.; Aaberg, Rosanne L.; Lopresti, Charles A.; Miley, Terri B.; Nichols, William E.; Strenge, Dennis L.

    2004-09-14

    This document contains detailed user instructions for a suite of utility codes developed for Rev. 1 of the Systems Assessment Capability. The suite of computer codes for Rev. 1 of Systems Assessment Capability performs many functions.

  18. Probabilistic structural analysis computer code (NESSUS)

    Science.gov (United States)

    Shiao, Michael C.

    1988-01-01

    Probabilistic structural analysis has been developed to analyze the effects of fluctuating loads, variable material properties, and uncertain analytical models especially for high performance structures such as SSME turbopump blades. The computer code NESSUS (Numerical Evaluation of Stochastic Structure Under Stress) was developed to serve as a primary computation tool for the characterization of the probabilistic structural response due to the stochastic environments by statistical description. The code consists of three major modules NESSUS/PRE, NESSUS/FEM, and NESSUS/FPI. NESSUS/PRE is a preprocessor which decomposes the spatially correlated random variables into a set of uncorrelated random variables using a modal analysis method. NESSUS/FEM is a finite element module which provides structural sensitivities to all the random variables considered. NESSUS/FPI is Fast Probability Integration method by which a cumulative distribution function or a probability density function is calculated.

  19. Code comparison for accelerator design and analysis

    International Nuclear Information System (INIS)

    This paper presents a comparison between results obtained from standard accelerator physics codes used for the design and analysis of sychrotrons and storage rings, with programs SYNCH, MAD, HARMON, PATRICIA, PATPET, BETA, DIMAD, MARYLIE and RACETRACK. In the analysis the authors have considered 5 (various size) lattices with large and small bend angles including AGS Booster (10 degrees bend) RHIC (2.24 degrees), SXLS, XLS (XUV ring with 45 degrees bend) and X-RAY rings. The differences in the integration methods used and the treatment of the fringe fields in these codes could lead to different results. The inclusion of nonlinear (e.g. dipole) terms may be necessary in these calculations specially for a small ring

  20. Code comparison for accelerator design and analysis

    International Nuclear Information System (INIS)

    We present a comparison between results obtained from standard accelerator physics codes used for the design and analysis of synchrotrons and storage rings, with programs SYNCH, MAD, HARMON, PATRICIA, PATPET, BETA, DIMAD, MARYLIE and RACE-TRACK. In our analysis we have considered 5 (various size) lattices with large and small angles including AGS Booster (10/degree/ bend), RHIC (2.24/degree/), SXLS, XLS (XUV ring with 45/degree/ bend) and X-RAY rings. The differences in the integration methods used and the treatment of the fringe fields in these codes could lead to different results. The inclusion of nonlinear (e.g., dipole) terms may be necessary in these calculations specially for a small ring. 12 refs., 6 figs., 10 tabs

  1. Development and validation of I-activation analysis code

    International Nuclear Information System (INIS)

    I-Activation Analysis Code (IAAC) is a nuclear depletion code which solves coupled Bateman equations for radioactive-transmutation and growth-decay system for large numbers of isotopes to get time evolution of decay products and nuclear activity. It is currently being developed primarily for neutron activation and radiation waste analysis, as a part of the code development activities. The code functions by separating long and short-lived isotopes and then uses the well-known matrix exponential method to quickly solve a large system of coupled, linear, first-order ordinary differential equations with constant coefficients for long-lived isotopes. This method allows a faster treatment of complex decay and transmutation schemes. The short-lived isotopes are solved using approximated decay-chain method. FENDL 3.0 neutron activation files are used for data library. Separate set of code modules are designed to read, decode, convert and condense the continuous-energy ACE formatted data into 175 VITAMIN-J energy groups. The new compiled library that includes half-lives and neutron absorption cross sections is then used as input source for nuclear data. The code is readily suitable for calculations pertaining to nuclear transmutation, activation and decay studies in mainly fusion applications and activation analyses. Details of the code and its primary validation performed for various test cases and material compositions, largely related to current ITER project specific neutronic and radiation analyses will be presented. The nuclear activity calculations are validated against FISPACT, available under EASY code system. (author)

  2. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  3. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  4. A Students Attendance System Using QR Code

    Directory of Open Access Journals (Sweden)

    Fadi Masalha

    2014-01-01

    Full Text Available Smartphones are becoming more preferred companions to users than desktops or notebooks. Knowing that smartphones are most popular with users at the age around 26, using smartphones to speed up the process of taking attendance by university instructors would save lecturing time and hence enhance the educational process. This paper proposes a system that is based on a QR code, which is being displayed for students during or at the beginning of each lecture. The students will need to scan the code in order to confirm their attendance. The paper explains the high level implementation details of the proposed system. It also discusses how the system verifies student identity to eliminate false registrations.

  5. Performance Analysis of Mobile WiMAX System using Turbo Coding Technique for Different Modulation Schemes under AWGN and Rayleigh Channel

    Directory of Open Access Journals (Sweden)

    P. Samundiswary

    2013-02-01

    Full Text Available In wireless communication, future demands must be met using more data throughput wireless technologies. Since bandwidth is limited and user demand continues to grow. This problem could be solved by WiMAX technology based on IEEE 802.16e specifications, which provides high data services on mobile environment. So, to support high data rate traffic, the error correction could be enhanced by incorporating a better channel coding technique in mobile WiMAX physical layer. Further, coding technique is used for providing reliable information through the transmission channel to the user. It is used to reduce the level of noise and interferences in electronic medium. The amount of error detection and correction required and its effectiveness depends on the signal to noise ratio. The advantages of Forward Error Correction (FEC are that a back-channel is not required, or that retransmission of data can often be avoided, at the cost of higher bandwidth requirements on average. In this paper, the performance of mobile WiMAX system with convolutional turbo coding is determined and analyzed for various modulation schemes under different channels. The BER performance of mobile WiMAX system using convolutional turbo Coding is determined and compared with the existing concatenated Reed Solomon(RS coding in the presence of AWGN and Rayleigh Channel. From the simulation results, it is verified that convolutional turbo coding provides better BER performance than concatenated RS coding

  6. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  7. EAI-oriented information classification code system in manufacturing enterprises

    Institute of Scientific and Technical Information of China (English)

    Junbiao WANG; Hu DENG; Jianjun JIANG; Binghong YANG; Bailing WANG

    2008-01-01

    Although the traditional information classifi-cation coding system in manufacturing enterprises (MEs) emphasizes the construction of code standards, it lacks the management of the code creation, code data transmission and so on. According to the demands of enterprise application integration (EAI) in manufacturing enter-prises, an enterprise application integration oriented information classification code system (EAIO-ICCS) is proposed. EAIO-ICCS expands the connotation of the information classification code system and assures the identity of the codes in manufacturing enterprises with unified management of codes at the view of its lifecycle.

  8. HELIAS module development for systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.

    2015-02-15

    In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

  9. Sequence Coding and Search System for licensee event reports: code listings. Volume 2

    International Nuclear Information System (INIS)

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. Volume 2 contains all valid and acceptable codes used for searching and encoding the LER data. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 2

  10. Automated Source Code Changes Classification for Effective Code Review and Analysis

    OpenAIRE

    Evgeny, G.

    2008-01-01

    Software development process is a complex sequence of actions having source code of working system as a result. All project participants should track changes in source code during work process to know what’s happening. However to make «manual» code review everyone should have corresponding technical skills and a lot of time to spend. This work describes usage of automated source code changes classification aimed to control source code evolution. The method bases on statistical clusterization ...

  11. Matlab Code for Structural Decomposition Analysis

    OpenAIRE

    Juan Tomas Sayago-Gomez

    2014-01-01

    This TechDoc describes the steps necessary to apply the Structural Decomposition Analysis (SDA) using Matlab. The code has two stages. The first stage, which comprises PrepSDA.m and RAS_SDA.m, prepares the data and the input required for SDA based on the accounting identities defined in Miller and Blair (2009) and Jackson and Schwarm (2011). The second stage (SDA.m) carries out the analysis and estimates the results based on the mathematical procedure in Yang and Lahr (2010) and Zhang and Lah...

  12. A Students Attendance System Using QR Code

    OpenAIRE

    Fadi Masalha; Nael Hirzallah

    2014-01-01

    Smartphones are becoming more preferred companions to users than desktops or notebooks. Knowing that smartphones are most popular with users at the age around 26, using smartphones to speed up the process of taking attendance by university instructors would save lecturing time and hence enhance the educational process. This paper proposes a system that is based on a QR code, which is being displayed for students during or at the beginning of each lecture. The students will need to scan the co...

  13. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

  14. Cooperative Regenerating Codes for Distributed Storage Systems

    OpenAIRE

    Shum, Kenneth W.

    2011-01-01

    When there are multiple node failures in a distributed storage system, regenerating the failed storage nodes individually in a one-by-one manner is suboptimal as far as repair-bandwidth minimization is concerned. If data exchange among the newcomers is enabled, we can get a better tradeoff between repair bandwidth and the storage per node. An explicit and optimal construction of cooperative regenerating code is illustrated.

  15. Safety analysis and the code development on radioactive waste disposal

    International Nuclear Information System (INIS)

    Regarding development of the safety analysis codes to be used for 'cross-check' (which is the evaluation of the validity of the safety analysis conducted by the licensee through cross comparison of the simulated result) of the sub-surface disposal conducted by the licensee, the codes are required to be capable of confirming the long term safety of the sub-surface disposal. The influence of the rainfall infiltration change on groundwater flow over the long term period due to climate change was studied. As a result, it was found that shoreline movement caused by the sea level change significantly influenced groundwater flow. Regarding development of the safety analysis codes to be used for 'cross-check' of the near surface disposal, it is important to efficiently simulate the groundwater flow with finely discretized mesh model. We therefore improved the memory allocation algorithm of the groundwater flow simulation code, TOUGH2 to be able to treat the large mesh model, such as several million cells. Modifications are made for the simulation support system, by adding the groundwater flow code 3D-SEEP which can treat land uplift and erosion and its associated modules. This modification not only improves efficiency but also allows to avoid human error. Moreover, sensitivity analysis of the unsaturated conditions such as infiltration rate on the migration of important nuclides of near surface disposal was conducted. As a result, influence of the unsaturated conditions on the exposed dose was evaluated. (author)

  16. SRAC95; general purpose neutronics code system

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).

  17. Modular ORIGEN-S for multi-physics code systems

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2011-07-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  18. Analysis of Void Fraction Distribution and Departure from Nucleate Boiling in Single Subchannel and Bundle Geometries Using Subchannel, System, and Computational Fluid Dynamics Codes

    Directory of Open Access Journals (Sweden)

    Taewan Kim

    2012-01-01

    Full Text Available In order to assess the accuracy and validity of subchannel, system, and computational fluid dynamics codes, the Paul Scherrer Institut has participated in the OECD/NRC PSBT benchmark with the thermal-hydraulic system code TRACE5.0 developed by US NRC, the subchannel code FLICA4 developed by CEA, and the computational fluid dynamic code STAR-CD developed by CD-adapco. The PSBT benchmark consists of a series of void distribution exercises and departure from nucleate boiling exercises. The results reveal that the prediction by the subchannel code FLICA4 agrees with the experimental data reasonably well in both steady-state and transient conditions. The analyses of single-subchannel experiments by means of the computational fluid dynamic code STAR-CD with the CD-adapco boiling model indicate that the prediction of the void fraction has no significant discrepancy from the experiments. The analyses with TRACE point out the necessity to perform additional assessment of the subcooled boiling model and bulk condensation model of TRACE.

  19. Blind Recognition Algorithm of Turbo Codes for Communication Intelligence Systems

    Directory of Open Access Journals (Sweden)

    Ali Naseri

    2011-11-01

    Full Text Available Turbo codes are widely used in land and space radio communication systems, and because of complexity of structure, are custom in military communication systems. In electronic warfare, COMINT systems make attempt to recognize codes by blind ways. In this Paper, the algorithm is proposed for blind recognition of turbo code parameters like code kind, code-word length, code rate, length of interleaver and delay blocks number of convolution code. The algorithm calculations volume is0.5L3+1.25L, therefore it is suitable for real time systems.

  20. Probabilistic analysis of crack containing structures with the PARIS code

    International Nuclear Information System (INIS)

    The basic features of the PARIS code which has been developed for the calculation of failure probabilities of crack containing structures are explained. An important issue in the reliability analysis of cracked components is the probabilistic leak-before-break behaviour. Formulae for the leak and break probabilities are derived and it is shown how a leak detection system influences the results. An example taken from nuclear applications illustrates the details of the probabilistic leak-before-break analysis. (orig.)

  1. Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

    International Nuclear Information System (INIS)

    Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes

  2. User's manual for ASTERIX-2: A two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)

  3. An Analysis Of Code Switching And Code Mixing Used In A Talk Show Hitam Putih

    OpenAIRE

    Sari, Dewi Maya

    2015-01-01

    In thesis entitled An Analysis of Code Switching and Code Mixing Used in Talk Show Hitam Putih, the author analyzes two types of code switching and code based on two types of mixed Wardaugh theory. The fourth type can be determined by the use of more than one language in an utterance. The purpose of this thesis is to find the types of code switching and code mixing contained in the speech Deddy Corbuzier as presenter in Talk Show Hitam Putih and Nadya Hutagalung as a celebrity guest. Steps ta...

  4. Suppression pool swell analysis using CFD code

    International Nuclear Information System (INIS)

    A two-dimensional axi-symmetric model of suppression pool of Containment Studies Facility (CSF) along with single vent pipe was modeled to estimate the jet and hydrodynamic loads due to flow of steam air mixture during simulated loss of coolant accident (LOCA). The analysis was carried out using CFD ACE+ software with Volume of Fluid (VOF) approach. The flow velocity variation through vent pipe was estimated using in-house containment thermal hydraulic code CONTRAN, was given as input at inlet boundary condition. The transient calculations were performed for 20 seconds and suppression pool level variation, pressure loads over the floor, walls and vent pipes etc were evaluated. (author)

  5. Thermal-hydraulic analysis for the LBE-cooled natural circulation reactor. Development of the MSG-COPD code and application to the system analysis. Research Document

    International Nuclear Information System (INIS)

    Thermal-hydraulic analysis for the Lead-Bismuth eutectic (LBE)-cooled natural circulation reactor has been conducted by using a combined plant dynamics code (MSG-COPD). MSG-COPD has been developed to consider the multi-dimensional thermal-hydraulics effect on the plant dynamics during transients. Plant dynamics analyses for the LBE-cooled STAR-LM reactor, which has been designed by Argonne National Laboratory in U.S.A., have been performed to understand the basic thermal-hydraulic characteristics of the natural circulation reactor. As a result, it has been made clear that cold coolant remains in the lower plenum by the thermal stratification in case of the ULOHS condition with a severe temperature gradient at the stratified surface in the lower plenum. In addition, the flow-redistribution effect in a core channels by the buoyancy force has been evaluated for a candidate LBE-cooled FBR plant concept (LBE-FR), which has been designed by JNC. A linear evaluation method for the flow-redistribution coefficient is proposed for the LBE-FR, and compared with the multi-dimensional results by MSG-COPD. In conclusion, the method shows sufficient performance for the prediction of the flow-redistribution coefficient for typical lateral power distributions in the core. (author)

  6. Blind Recognition Algorithm of Turbo Codes for Communication Intelligence Systems

    OpenAIRE

    Ali Naseri; Omid Azmoon; Samad Fazeli

    2011-01-01

    Turbo codes are widely used in land and space radio communication systems, and because of complexity of structure, are custom in military communication systems. In electronic warfare, COMINT systems make attempt to recognize codes by blind ways. In this Paper, the algorithm is proposed for blind recognition of turbo code parameters like code kind, code-word length, code rate, length of interleaver and delay blocks number of convolution code. The algorithm calculations volume is0.5L3+1.25L, th...

  7. Problems of optimal data coding in hodoscopic systems

    International Nuclear Information System (INIS)

    An analogy system of algebraic coding theory and of hodoscopic system coding theory is considered. The connection between main parameters of coding devices and parameters of parallel coders applied in hodoscopic systems is established. The efficiency of using a proposed analogy system is illustrated on some examples of designing parallel coders with given properties

  8. Upgrades to the NESS (Nuclear Engine System Simulation) Code

    Science.gov (United States)

    Fittje, James E.

    2007-01-01

    In support of the President's Vision for Space Exploration, the Nuclear Thermal Rocket (NTR) concept is being evaluated as a potential propulsion technology for human expeditions to the moon and Mars. The need for exceptional propulsion system performance in these missions has been documented in numerous studies, and was the primary focus of a considerable effort undertaken during the 1960's and 1970's. The NASA Glenn Research Center is leveraging this past NTR investment in their vehicle concepts and mission analysis studies with the aid of the Nuclear Engine System Simulation (NESS) code. This paper presents the additional capabilities and upgrades made to this code in order to perform higher fidelity NTR propulsion system analysis and design.

  9. Analysis of SMA hybrid composite structures using commercial codes

    Science.gov (United States)

    Turner, Travis L.; Patel, Hemant D.

    2004-07-01

    A thermomechanical model for shape memory alloy (SMA) actuators and SMA hybrid composite (SMAHC) structures has been recently implemented in the commercial finite element codes MSC.Nastran and ABAQUS. The model may be easily implemented in any code that has the capability for analysis of laminated composite structures with temperature dependent material properties. The model is also relatively easy to use and requires input of only fundamental engineering properties. A brief description of the model is presented, followed by discussion of implementation and usage in the commercial codes. Results are presented from static and dynamic analysis of SMAHC beams of two types; a beam clamped at each end and a cantilevered beam. Nonlinear static (post-buckling) and random response analyses are demonstrated for the first specimen. Static deflection (shape) control is demonstrated for the cantilevered beam. Approaches for modeling SMAHC material systems with embedded SMA in ribbon and small round wire product forms are demonstrated and compared. The results from the commercial codes are compared to those from a research code as validation of the commercial implementations; excellent correlation is achieved in all cases.

  10. Digital Image Analysis for Detechip Code Determination

    Directory of Open Access Journals (Sweden)

    Marcus Lyon

    2012-09-01

    Full Text Available DETECHIP® is a molecular sensing array used for identification of a large variety of substances. Previous methodology for the analysis of DETECHIP® used human vision to distinguish color changes induced by the presence of the analyte of interest. This paper describes several analysis techniques using digital images of DETECHIP®. Both a digital camera and flatbed desktop photo scanner were used to obtain Jpeg images. Color information within these digital images was obtained through the measurement of redgreen-blue (RGB values using software such as GIMP, Photoshop and ImageJ. Several different techniques were used to evaluate these color changes. It was determined that the flatbed scanner produced in the clearest and more reproducible images. Furthermore, codes obtained using a macro written for use within ImageJ showed improved consistency versus pervious methods.

  11. Digital Image Analysis for Detechip Code Determination

    Directory of Open Access Journals (Sweden)

    Marcus Lyon

    2012-08-01

    Full Text Available DETECHIP® is a molecular sensing array used for identification of a large variety of substances. Previous methodology for the analysis of DETECHIP® used human vision to distinguish color changes induced by the presence of the analyte of interest. This paper describes several analysis techniques using digital images of DETECHIP® . Both a digital camera and flatbed desktop photo scanner were used to obtain Jpeg images. Color information within these digital images was obtained through the measurement of redgreen-blue (RGB values using software such as GIMP, Photoshop and ImageJ. Several different techniques were used to evaluate these color changes. It was determined that the flatbed scanner produced in the clearest and more reproducible images. Furthermore, codes obtained using a macro written for use within ImageJ showed improved consistency versus pervious methods.

  12. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)

  13. Performance Analysis of UWB System Based on Turbo Coding%基于Turbo编码的超宽带系统性能分析

    Institute of Scientific and Technical Information of China (English)

    陈煌林

    2011-01-01

    In order to decrease severe time dispersion, a Turbo channel coding introduced into UWB system is put forward and the bit error rates performance of UWB system based on Turbo coding is analyzed and simulated in different wireless indoor environment. The indoor wireless environment is modeled as a modified Saleh and Valenzuela (SV) channel which is put forward by IEEE802. 15. 3a. To decrease complexity of the iterative decode,LOG-MAP algorithm is adopted. The simulation results show that comparing with the system without coding, UWB system with Turbo coding offers considerable coding gain. It demonstrates that the performance of the UWB system can be substantially improved by increasing the number of iteration.%为了降低严重的时间弥散影响,提出了一种Turbo信道编码方案引入超宽带系统中,分析和仿真了在不同无线室内环境下基于Turbo编码的超宽带系统的误比特率性能.无线室内环境是由IEEE 802.15.3a提出的修正的SV信道模型.为了降低迭代译码的复杂度,采用了LOG-MAP算法.仿真结果表明,相比于无编码的系统,具有Turbo编码的超宽带系统在不同无线室内环境下提供了可观的编码增益,随着迭代次数的增加,超宽带系统的性能得到了改善.

  14. MTR coded PRML systems for perpendicular magnetic recording

    Energy Technology Data Exchange (ETDEWEB)

    Okamoto, Yoshihiro E-mail: okamoto@rec.ee.ehime-u.ac.jp; Sato, Mitsuteru; Osawa, Hisashi; Saito, Hidetoshi; Muraoka, Hiroaki; Nakamura, Yoshihisa

    2001-10-01

    We evaluate the BER performance of various MTR coded PRML systems characterized by the polynomials with only positive coefficients in a perpendicular magnetic recording channel using a double-layered medium with jitter-like noise by computer simulation. The results show that ((3)/(4)) MTR coded PRML systems exhibit good performances compared with ((16)/(17)) MTR coded PRML systems.

  15. DESIGN OF EXACT REGENERATING HIERARCHICAL CODE FOR DISTRIBUTED STORAGE SYSTEM

    Institute of Scientific and Technical Information of China (English)

    Hao Jie; Lu Yanbo; Liu Xinji; Xia Shutao

    2013-01-01

    Erasure code is widely used as the redundancy scheme in distributed storage system.When a storage node fails,the repair process often requires to transfer a large amount of data.Regenerating code and hierarchical code are two classes of codes proposed to reduce the repair bandwidth cost.Regenerating codes reduce the amount of data transferred by each helping node,while hierarchical codes reduce the number of nodes participating in the repair process.In this paper,we propose a "sub-code nesting framework" to combine them together.The resulting regenerating hierarchical code has low repair degree as hierarchical code and lower repair cost than hierarchical code.Our code can achieve exact regeneration of the failed node,and has the additional property of low updating complexity.

  16. Development and assessment of best estimate integrated safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

    2007-03-15

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

  17. Development and assessment of best estimate integrated safety analysis code

    International Nuclear Information System (INIS)

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published

  18. Development of Tritium Permeation Analysis Code and Tritium Transport in a High Temperature Gas-Cooled Reactor Coupled with Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2010-06-01

    Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.

  19. DEPTH-CHARGE static and time-dependent perturbation/sensitivity system for nuclear reactor core analysis. Revision I. [DEPTH-CHARGE code

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1985-04-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code black for both static and time-dependent perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Laboratory. The DEPTH module (coupled with VENTURE) solves for the three adjoint functions of Depletion Perturbation Theory and calculates the desired time-dependent derivatives of the response with respect to the nuclide concentrations and nuclear data utilized in the reference model. The CHARGE code is a collection of utility routines for general data manipulation and input preparation and considerably extends the usefulness of the system through the automatic generation of adjoint sources, estimated perturbed responses, and relative data sensitivity coefficients. Combined, the DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analyses of realistic multidimensional reactor models. This current documentation incorporates minor revisions to the original DEPTH-CHARGE documentation (ORNL/CSD-78) to reflect some new capabilities within the individual codes.

  20. Analysis of the nodalization effect on the simulation of atmospheric stratification in the ThAI TH13 experiment using the containment code system COCOSYS

    International Nuclear Information System (INIS)

    During a severe accident hydrogen generation in the core tat temperatures above 1000 C is possible due to oxidation reaction between water vapor and the zirconium in the fuel element cladding. In the frame of the international standard problem experiment ISP-47 step2 the experiment ThAI TH13 on the hydrogen distribution was performed for a containment model and calculated using CFD and lumped parameter (LP) codes. The experiment uses helium instead of the explosive hydrogen. The paper is focused on the analysis of the nodalization effect on the simulation of the atmospheric stratification using the containment analysis code COCOSYS. The simulation of flow phenomena like stratified layer decomposition and homogenization within the containment is dependent on the vertical subdivision of the containment nodalization.

  1. Development and verification of a thermo-hydraulic simulation code for systems transient in 'Monju' (COPD code)

    International Nuclear Information System (INIS)

    Large system simulation codes are needed for design and safety analysis. A thermal-hydraulic simulation code for systems transient in ''Monju'' (COPD code) was developed and verified with experimental data from an experimental LMFBR ''Joyo'', 50 MWt steam generator test facility and scaled test sections of reactor vessel plenum. This paper summarizes numerical models of this code and their verifications with experimental data. Especially, a simplified analytical model to predict the transient behavior in a reactor vessel plenum is presented in detail, since this behavior has an important effect that must be taken into account in a plant thermal transient, while the reactor is tripped. The COPD is applied to design and safety analysis in ''Monju'' as follows ; (1) Safety analysis with regard to core cooling in anticipated incidents. (2) Plant thermo-hydraulic analysis for setting the design condition in thermal stress analysis and evaluation of components and pipings. (3) Control performance analysis on plant operation for design and evaluation of plant control system. Each of the above analyses requires different predictions of plant response to be analyzed. Therefore, appropriate models and input data are used in the design and evaluation according to the purpose of the analysis. This code was developed and verified under a contract with PNC. (author)

  2. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor

    International Nuclear Information System (INIS)

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

  3. A preliminary uncertainty analysis of phenomenological inputs in TEXAS-V code

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. H.; Kim, H. D.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Uncertainty analysis is important step in safety analysis of nuclear power plants. The better estimate for the computer codes is on the increase instead of conservative codes. These efforts aim to get more precise evaluation of safety margins, and aim at determining the rate of change in the prediction of codes with one or more input parameters varies within its range of interest. From this point of view, a severe accident uncertainty analysis system, SAUNA, has been improved for TEXAS-V FCI uncertainty analysis. The main objective of this paper is to present the TEXAS FCI uncertainty analysis results implemented through the SAUNA code

  4. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)

  5. Advanced analysis of the CEA-NEA/OECD WWER-1000 coolant transient benchmark with the coupled system code ATHLET/BIPR-WWER

    International Nuclear Information System (INIS)

    Recent studies showed the necessity of a detailed modelling of the core outlet region of the WWER-1000 reactor where the thermocouples are located. Solving of this problem is of primary importance for the validation of the coupled system code ATHLET/BIPR-WWER on local parameters. Therefore, special attention is paid on the reactor pressure vessel model and its nodalization schema and in particular the fluid mixing phenomena at assemblies' outlets. For this purpose additional thermal-hydraulic channels modelling the flow along the guide tubes are introduced in the reactor core. With the new advanced modelling again the benchmark problems of Phase 1 of the CEA-NEA/OECD WWER-1000 Coolant Transient Benchmark are analysed. On the base of data comparison with the experimental measurements (Phase 2, Exercise 1) the mixing phenomena at assembly head is estimated and mixing coefficients are introduced in the thermal-hydraulic core outlet models of the coupled system code ATHLET/BIPR-WWER (Authors)

  6. On Analyzing LDPC Codes over Multiantenna MC-CDMA System

    Directory of Open Access Journals (Sweden)

    S. Suresh Kumar

    2014-01-01

    Full Text Available Multiantenna multicarrier code-division multiple access (MC-CDMA technique has been attracting much attention for designing future broadband wireless systems. In addition, low-density parity-check (LDPC code, a promising near-optimal error correction code, is also being widely considered in next generation communication systems. In this paper, we propose a simple method to construct a regular quasicyclic low-density parity-check (QC-LDPC code to improve the transmission performance over the precoded MC-CDMA system with limited feedback. Simulation results show that the coding gain of the proposed QC-LDPC codes is larger than that of the Reed-Solomon codes, and the performance of the multiantenna MC-CDMA system can be greatly improved by these QC-LDPC codes when the data rate is high.

  7. Improved FEC Code Based on Concatenated Code for Optical Transmission Systems

    Institute of Scientific and Technical Information of China (English)

    YUAN Jian-guo; JIANG Ze; MAO You-ju

    2006-01-01

    The improved three novel schemes of the super forward error correction(super-FEC) concatenated codes are proposed after the development trend of long-haul optical transmission systems and the defects of the existing FEC codes have been analyzed. The performance simulation of the Reed-Solomon(RS)+Bose-Chaudhuri-Hocguenghem(BCH) inner-outer serial concatenated code is implemented and the conceptions of encoding/decoding the parallel-concatenated code are presented. Furthermore,the simulation results for the RS(255,239)+RS(255,239) code and the RS(255,239)+RS(255,223) code show that the two consecutive concatenated codes are a superior coding scheme with such advantages as the better error correction,moderate redundancy and easy realization compared to the classic RS(255,239) code and other codes,and their signal to noise ratio gains are respectively 2~3 dB more than that of the RS(255,239)code at the bit error rate of 1×10-13. Finally,the frame structure of the novel consecutive concatenated code is arranged to lay a firm foundation in designing its hardware.

  8. Simulation and analysis of void drift using sub-channel analysis code and CFD code

    Energy Technology Data Exchange (ETDEWEB)

    Pang, Bo; Cheng, Xu; Otic, Ivan [Karlsruhe Institute of Technology (KIT) (Germany). Inst. of Fusion and Reactor Technology (IFRT)

    2012-11-01

    Prediction accuracy of a sub-channel analysis depends strongly on the modeling of the interchannel transverse exchange effect. Disregarding the forced mixing effects caused by extra constructive elements the natural inter-channel transverse exchange effect can be decomposed into [1] [2] [3]: turbulent mixing (TM) due to the natural eddy diffusion, diversion cross flow (DC) induced by radial pressure gradient and void drift (VD) specially under two-phase flow conditions. Among the three components, the physical mechanism of void drift is not well clarified. Previous to the time and cost demanding experimental research a systematic numerical simulation of the inter-channel exchange effect with CFD code can provide supplemental information about the physical mechanism behind the not well clarified void drift phenomena. Compared to sub-channel analysis code, CFD code solves the flow dynamic problem with a much finer mesh and in a more physical way. The inter-channel exchange terms are solved in the conservation equations rather than modeled with closure equations. Furthermore, the inter-phase exchange terms are also taken into account. A better understanding of the void drift phenomenon and a modification of the void drift models in a sub-channel analysis code basing on the CFD analysis can be achieved. In present study, both sub-channel and CFD analysis are carried out for studying the void drift in a rod bundle geometry. A model is proposed to determine the sub-channel scale void drift mass flux based on the CFD simulation results. (orig.)

  9. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  10. Analysis and Design of Tuned Turbo Codes

    CERN Document Server

    Koller, Christian; Kliewer, Joerg; Vatta, Francesca; Zigangirov, Kamil S; Costello, Daniel J

    2010-01-01

    It has been widely observed that there exists a fundamental trade-off between the minimum distance properties and the iterative decoding convergence behavior of turbo-like codes. While capacity achieving code ensembles typically are asymptotically bad in the sense that their minimum distance does not grow linearly with block length, and they therefore exhibit an error floor at moderate-to-high signal to noise ratios, asymptotically good codes usually converge further away from channel capacity. In this paper, we introduce the concept of tuned turbo codes, a family of asymptotically good hybrid concatenated code ensembles, where minimum distance growth rates, convergence thresholds, and code rates can be traded-off using two tuning parameters, {\\lambda} and {\\mu}. By decreasing {\\lambda}, the asymptotic minimum distance growth rate is reduced for the sake of improved iterative decoding convergence behavior, while increasing {\\lambda} raises the growth rate at the expense of worse convergence behavior, and thus...

  11. Validation of OPERA3D PCMI Analysis Code

    International Nuclear Information System (INIS)

    This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel

  12. Diffuser augmented wind turbine analysis code

    Science.gov (United States)

    Carroll, Jonathan

    Wind Energy is becoming a significant source of energy throughout the world. This ever increasing field will potentially reach the limit of availability and practicality with the wind farm sites and size of the turbine itself. Therefore, it is necessary to develop innovative wind capturing devices that can produce energy in the locations where large conventional horizontal axis wind turbines (HAWTs) are too impractical to install and operate. A diffuser augmented wind turbine (DAWT) is one such innovation. DAWTs increase the power output of the rotor by increasing the wind speed into the rotor using a duct. Currently, developing these turbines is an involved process using time consuming Computational Fluid Dynamics codes. A simple and quick design tool is necessary for designers to develop efficient energy capturing devices. This work lays out the theory for a quick analysis tool for DAWTs using an axisymmetric surface vorticity method. This method allows for quick analysis of duct, hubs and rotors giving designers a general idea of the power output of the proposed hub, blade and duct geometry. The method would be similar to the way blade element momentum theory is used to design conventional HAWTs. It is determined that the presented method is viable for preliminary design of DAWTs.

  13. The analysis of convolutional codes via the extended Smith algorithm

    Science.gov (United States)

    Mceliece, R. J.; Onyszchuk, I.

    1993-01-01

    Convolutional codes have been the central part of most error-control systems in deep-space communication for many years. Almost all such applications, however, have used the restricted class of (n,1), also known as 'rate 1/n,' convolutional codes. The more general class of (n,k) convolutional codes contains many potentially useful codes, but their algebraic theory is difficult and has proved to be a stumbling block in the evolution of convolutional coding systems. In this article, the situation is improved by describing a set of practical algorithms for computing certain basic things about a convolutional code (among them the degree, the Forney indices, a minimal generator matrix, and a parity-check matrix), which are usually needed before a system using the code can be built. The approach is based on the classic Forney theory for convolutional codes, together with the extended Smith algorithm for polynomial matrices, which is introduced in this article.

  14. A Content Analysis of Student Conduct Codes

    OpenAIRE

    Martin, Janice Earlene

    2004-01-01

    Scholars in the field of student judicial affairs have recommended that institutions remove all legal terminology and references in student conduct codes and create codes based on student development theory and practice (Dannells, 1997; Gehring, 2001; Stoner & Cerminara 1990; Stoner, 2000). The purpose of this study was to analyze student conduct codes to determine the extent to which college and university administrators have adopted Stoner and Cerminara, Gehring, and Pavela's suggestions. ...

  15. CORRELATING FEATURES AND CODE BY DYNAMIC AND SEMANTIC ANALYSIS

    Directory of Open Access Journals (Sweden)

    Ren Wu

    2015-10-01

    Full Text Available One major problem in maintaining a software system is to understand how many functional features in the system and how these features are implemented. In this paper a novel approach for locating features in code by semantic and dynamic analysis is proposed. The method process consists of three steps: The first uses the execution traces as text corpus and the method calls involved in the traces as terms of document. The second ranks the method calls in order to filter out omnipresent methods by setting a threshold. And the third step treats feature-traces as first class entities and extracts identifiers from the rest method source code and a trace-by-identifier matrix is generated. Then a semantic analysis model-LDA is applied on the matrix to extract topics, which act as functional features. Through building several corresponding matrices, the relations between features and code can be obtained for comprehending the system functional intents. A case study is presented and the execution results of this approach can be used to guide future research.

  16. Performance Analysis of Wavelet Channel Coding in COST207-based Channel Models on Simulated Radio-over-Fiber Systems at the W-Band

    DEFF Research Database (Denmark)

    Cavalcante, Lucas Costa Pereira; Silveira, Luiz F. Q.; Rommel, Simon;

    2016-01-01

    , such systems use diversity schemes in combination with digital signal processing (DSP) techniques to overcome effects such as fading and inter-symbol interference (ISI). Wavelet Channel Coding (WCC) has emerged as a technique to minimize the fading effects of wireless channels, which is a mayor challenge...... in systems operating in the millimeter wave regime. This work takes the WCC one step beyond by performance evaluation in terms of bit error probability, over time-varying, frequency-selective multipath Rayleigh fading channels. The adopted propagation model follows the COST207 norm, the main international...

  17. Study on the entire system of maintenance codes and standards

    International Nuclear Information System (INIS)

    In this study, a structure of code and standard system for plant maintenance is discussed along a process of maintenance activities. As a result of consideration, it was concluded as follows. (1) It is assumed that the entire system of maintenance codes and standards consists of four standards, that is, standards regarding maintenance planning, maintenance implementation, evaluation of inspection/maintenance results and corrective measures. (2) The maintenance guidelines and fitness-for-service codes discussed already so far occupies a position in the entire system of maintenance codes and standards. (3) Maintenance codes and standards, which have higher priority, should be developed. (author)

  18. Simplified computer codes for cask impact analysis

    International Nuclear Information System (INIS)

    In regard to the evaluation of the acceleration and deformation of casks, the simplified computer codes make analyses economical and decrease input and calculation time. The results obtained by the simplified computer codes have enough adequacy for their practical use. (J.P.N.)

  19. A New Arithmetic Coding System Combining Source Channel Coding and MAP Decoding

    Institute of Scientific and Technical Information of China (English)

    PANG Yu-ye; SUN Jun; WANG Jia

    2007-01-01

    A new arithmetic coding system combining source channel coding and maximum a posteriori decoding were proposed.It combines source coding and error correction tasks into one unified process by introducing an adaptive forbidden symbol.The proposed system achieves fixed length code words by adaptively adjusting the probability of the forbidden symbol and adding tail digits of variable length.The corresponding improved MAP decoding metric was derived.The proposed system can improve the performance.Simulations were performed on AWGN channels with various noise levels by using both hard and soft decision with BPSK modulation.The results show its performance is slightly better than that of our adaptive arithmetic error correcting coding system using a forbidden symbol.

  20. Communication Systems Simulator with Error Correcting Codes Using MATLAB

    Science.gov (United States)

    Gomez, C.; Gonzalez, J. E.; Pardo, J. M.

    2003-01-01

    In this work, the characteristics of a simulator for channel coding techniques used in communication systems, are described. This software has been designed for engineering students in order to facilitate the understanding of how the error correcting codes work. To help students understand easily the concepts related to these kinds of codes, a…

  1. Deductive Glue Code Synthesis for Embedded Software Systems Based on Code Patterns

    Science.gov (United States)

    Liu, Jian; Fu, Jicheng; Zhang, Yansheng; Bastani, Farokh; Yen, I-Ling; Tai, Ann; Chau, Savio N.

    2006-01-01

    Automated code synthesis is a constructive process that can be used to generate programs from specifications. It can, thus, greatly reduce the software development cost and time. The use of formal code synthesis approach for software generation further increases the dependability of the system. Though code synthesis has many potential benefits, the synthesis techniques are still limited. Meanwhile, components are widely used in embedded system development. Applying code synthesis to component based software development (CBSD) process can greatly enhance the capability of code synthesis while reducing the component composition efforts. In this paper, we discuss the issues and techniques for applying deductive code synthesis techniques to CBSD. For deductive synthesis in CBSD, a rule base is the key for inferring appropriate component composition. We use the code patterns to guide the development of rules. Code patterns have been proposed to capture the typical usages of the components. Several general composition operations have been identified to facilitate systematic composition. We present the technique for rule development and automated generation of new patterns from existing code patterns. A case study of using this method in building a real-time control system is also presented.

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

    International Nuclear Information System (INIS)

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

  3. Parallelization of Subchannel Analysis Code MATRA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A stand-alone calculation of MATRA code used up pertinent computing time for the thermal margin calculations while a relatively considerable time is needed to solve the whole core pin-by-pin problems. In addition, it is strongly required to improve the computation speed of the MATRA code to satisfy the overall performance of the multi-physics coupling calculations. Therefore, a parallel approach to improve and optimize the computability of the MATRA code is proposed and verified in this study. The parallel algorithm is embodied in the MATRA code using the MPI communication method and the modification of the previous code structure was minimized. An improvement is confirmed by comparing the results between the single and multiple processor algorithms. The speedup and efficiency are also evaluated when increasing the number of processors. The parallel algorithm was implemented to the subchannel code MATRA using the MPI. The performance of the parallel algorithm was verified by comparing the results with those from the MATRA with the single processor. It is also noticed that the performance of the MATRA code was greatly improved by implementing the parallel algorithm for the 1/8 core and whole core problems.

  4. A Coding System for Qualitative Studies of the Information-Seeking Process in Computer Science Research

    Science.gov (United States)

    Moral, Cristian; de Antonio, Angelica; Ferre, Xavier; Lara, Graciela

    2015-01-01

    Introduction: In this article we propose a qualitative analysis tool--a coding system--that can support the formalisation of the information-seeking process in a specific field: research in computer science. Method: In order to elaborate the coding system, we have conducted a set of qualitative studies, more specifically a focus group and some…

  5. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  6. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  7. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  8. ANACROM - A computer code for chromatogram analysis

    International Nuclear Information System (INIS)

    The computer code was developed for automatic research of peaks and evaluation of chromatogram parameters as : center, height, area, medium - height width (FWHM) and the rate FWHM/center of each peak. (Author)

  9. 航空通信系统中信道编码理论及应用分析%Theoretical Analysis and Application of Channel Coding in Aviation Communication Systems

    Institute of Scientific and Technical Information of China (English)

    邹星; 李金喜; 丁勇飞; 方正

    2015-01-01

    Generally, the aviation communication system has some constraints, such as power, bandwidth and latency. To satisfy its requirements of high coding gain and low latency to channel coding, the paper derives the theoretical Shannon limit on the minimum SNR required for coded system, especially for the binary communication system. And then, the performance of several typical kinds of channel coding is analyzed through simulation. Finally, qualitative analysis is made to the application scope of the different channel coding methods with consideration of the technical indicators of them. The conclusion has engineering significance for the selection of channel coding in aviation communication systems.%一般地,航空通信系统具有功率受限、带宽受限、时延受限的特点。针对航空通信系统对信道编码高增益和低时延的要求,首先理论推导通信系统(特别是二进制通信系统)应用r码率信道编码获得编码增益的香农极限值,并仿真对比分析几类典型信道编码的编码性能,最后结合工程实现考虑的技术指标对几类信道编码的应用范围进行了定性分析,其对航空通信系统的信道编码选取具有工程指导意义。

  10. Preliminary Coupling of MATRA Code for Multi-physics Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described.

  11. 滴眼药液监码系统的功能分析与研究%Functional Analysis and Research on the System of Monitoring Code of Eye Drops

    Institute of Scientific and Technical Information of China (English)

    王广林

    2015-01-01

    According to the "drug regulatory code production line Fu code system" requirements formulated by the State Food and drug administration, combined with Shapuaisi eye drops of actual production and management requirements, to meet the ba-sic needs of, the function of eye drop physic liquor supervision code system were analysis and research, describes the project background, research significance and system function, and discusses the design and implementation of monitoring system func-tion code.%根据国家食品药品监督管理局制定的对“药品监管码生产线赋码系统”的需求,结合莎普爱思滴眼液的实际生产情况及管理要求,在满足基本需求的基础上,对滴眼药液监码系统功能进行了分析与研究,阐述了项目的背景、研究意义以及系统的功能,并探讨了监码系统功能的设计与实现。

  12. Security Concerns and Countermeasures in Network Coding Based Communications Systems

    DEFF Research Database (Denmark)

    Talooki, Vahid; Bassoli, Riccardo; Roetter, Daniel Enrique Lucani;

    2015-01-01

    This survey paper shows the state of the art in security mechanisms, where a deep review of the current research and the status of this topic is carried out. We start by introducing network coding and its variety applications in enhancing current traditional networks. In particular, we analyze two...... key protocol types, namely, state-aware and stateless protocols, specifying the benefits and disadvantages of each one of them. We also present the key security assumptions of network coding (NC) systems as well as a detailed analysis of the security goals and threats, both passive and active....... This paper also presents a detailed taxonomy and a timeline of the different NC security mechanisms and schemes reported in the literature. Current proposed security mechanisms and schemes for NC in the literature are classified later. Finally a timeline of these mechanism and schemes is presented....

  13. Morse Monte Carlo Radiation Transport Code System

    Energy Technology Data Exchange (ETDEWEB)

    Emmett, M.B.

    1975-02-01

    The report contains sections containing descriptions of the MORSE and PICTURE codes, input descriptions, sample problems, deviations of the physical equations and explanations of the various error messages. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. The PICTURE code provide aid in preparing correct input data for the combinatorial geometry package CG. It provides a printed view of arbitrary two-dimensional slices through the geometry. By inspecting these pictures one may determine if the geometry specified by the input cards is indeed the desired geometry. 23 refs. (WRF)

  14. Analysis on Application of Turbo Product Code in UAV TT&C System%Turbo乘积码在无人机测控技术中的应用分析

    Institute of Scientific and Technical Information of China (English)

    周侃; 金松坡

    2012-01-01

    Turbo乘积码是一种易于硬件实现的分组码,具有延时短和纠错性能好等优越性。通过对Turbo乘积码Chase软输出改进算法的分析和仿真,得出了不同码率和测试序列个数等参数对Turbo乘积码译码性能的影响。结合无人机常用的BPSK测控信号进行了仿真实验,对仿真结果进行了性能分析,验证了Turbo乘积码在无人机测控系统中应用的可行性,并给出了Turbo乘积码在无人机测控领域应用的建议参数。%Turbo product code,as a block code easy to be implemented in hardware,has the advantages of short delay and good error correction performance.By an analysis and simulation of the improved Chase soft-output algorithm for Turbo product code,the effects of parameters such as bit rate and number of test sequences on the performance of Turbo product code decoding are obtained.Simulation experiments are conducted with the BPSK signal typically used by UAV.The analysis of the simulation results verifies the feasibility of applying the Turbo product code in UAV TTC system.Parameters are also proposed for the application of Turbo product code in UAV TTC system.

  15. PIPE STRESS and VERPIP codes for stress analysis and verifications of PEC reactor piping

    International Nuclear Information System (INIS)

    To design LMFBR piping systems following ASME Sct. III requirements unusual flexibility computer codes are to be adopted to consider piping and its guard-tube. For this purpose PIPE STRESS code previously prepared by Southern-Service, has been modified. Some subroutine for detailed stress analysis and principal stress calculations on all the sections of piping have been written and fitted in the code. Plotter can also be used. VERPIP code for automatic verifications of piping as class 1 Sct. III prescriptions has been also prepared. The results of PIPE STRESS and VERPIP codes application to PEC piping are in section III of this report

  16. A Semantic Analysis Method for Scientific and Engineering Code

    Science.gov (United States)

    Stewart, Mark E. M.

    1998-01-01

    This paper develops a procedure to statically analyze aspects of the meaning or semantics of scientific and engineering code. The analysis involves adding semantic declarations to a user's code and parsing this semantic knowledge with the original code using multiple expert parsers. These semantic parsers are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. In practice, a user would submit code with semantic declarations of primitive variables to the analysis procedure, and its semantic parsers would automatically recognize and document some static, semantic concepts and locate some program semantic errors. A prototype implementation of this analysis procedure is demonstrated. Further, the relationship between the fundamental algebraic manipulations of equations and the parsing of expressions is explained. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.

  17. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  18. Development status of Severe Accident Analysis Code SAMPSON

    Energy Technology Data Exchange (ETDEWEB)

    Iwashita, Tsuyoshi; Ujita, Hiroshi [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan)

    2000-11-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)

  19. AVS 3D Video Coding Technology and System

    Institute of Scientific and Technical Information of China (English)

    Siwei Ma; Shiqi Wang; Wen Gao

    2012-01-01

    Following the success of the audio video standard (AVS) for 2D video coding, in 2008, the China AVS workgroup started developing 3D video (3DV) coding techniques. In this paper, we discuss the background, technical features, and applications of AVS 3DV coding technology. We introduce two core techniques used in AVS 3DV coding: inter-view prediction and enhanced stereo packing coding. We elaborate on these techniques, which are used in the AVS real-time 3DV encoder. An application of the AVS 3DV coding system is presented to show the great practical value of this system. Simulation results show that the advanced techniques used in AVS 3DV coding provide remarkable coding gain compared with techniques used in a simulcast scheme.

  20. Generalized optical code construction for enhanced and Modified Double Weight like codes without mapping for SAC-OCDMA systems

    Science.gov (United States)

    Kumawat, Soma; Ravi Kumar, M.

    2016-07-01

    Double Weight (DW) code family is one of the coding schemes proposed for Spectral Amplitude Coding-Optical Code Division Multiple Access (SAC-OCDMA) systems. Modified Double Weight (MDW) code for even weights and Enhanced Double Weight (EDW) code for odd weights are two algorithms extending the use of DW code for SAC-OCDMA systems. The above mentioned codes use mapping technique to provide codes for higher number of users. A new generalized algorithm to construct EDW and MDW like codes without mapping for any weight greater than 2 is proposed. A single code construction algorithm gives same length increment, Bit Error Rate (BER) calculation and other properties for all weights greater than 2. Algorithm first constructs a generalized basic matrix which is repeated in a different way to produce the codes for all users (different from mapping). The generalized code is analysed for BER using balanced detection and direct detection techniques.

  1. Deductive Evaluation: Formal Code Analysis With Low User Burden

    Science.gov (United States)

    Di Vito, Ben. L

    2016-01-01

    We describe a framework for symbolically evaluating iterative C code using a deductive approach that automatically discovers and proves program properties. Although verification is not performed, the method can infer detailed program behavior. Software engineering work flows could be enhanced by this type of analysis. Floyd-Hoare verification principles are applied to synthesize loop invariants, using a library of iteration-specific deductive knowledge. When needed, theorem proving is interleaved with evaluation and performed on the fly. Evaluation results take the form of inferred expressions and type constraints for values of program variables. An implementation using PVS (Prototype Verification System) is presented along with results for sample C functions.

  2. Development of analysis code of immersed decay heat removal system for fast reactor%快堆浸入式事故余热排出系统程序开发

    Institute of Scientific and Technical Information of China (English)

    钱鸿涛; 李政昕; 胡文军; 宫宇

    2015-01-01

    To meet the need of demonstration fast reactor design,a thermal-hydraulic analysis code of immersed decay heat remov-al system was developed based on a French fast reactor system code OASIS,with the introduction of the thermal stratification and inter-wrapper flow models.An integrated model was developed for the decay heat removal system of CEFR using the code,and the performances at steady and transient station black-out were analyzed.The calculation results were validated with other codes.%针对示范快堆的设计需要,在法国快堆系统程序 OASIS 的基础上,引入热分层与盒间流模型,开发了浸入式事故余热排出系统分析程序。利用该程序对 CEFR 的非能动事故余热排出系统进行了整体建模,分析了稳态和全厂断电工况下的性能,并利用其他系统程序的结果进行了验证。结果表明:该程序能较好地反映事故余热排出系统的瞬态变化过程。

  3. Evaluation of the DRAGON code for VHTR design analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  4. Coding Across Multicodes and Time in CDMA Systems Employing MMSE Multiuser Detector

    Directory of Open Access Journals (Sweden)

    Park Jeongsoon

    2004-01-01

    Full Text Available When combining a multicode CDMA system with convolutional coding, two methods have been considered in the literature. In one method, coding is across time in each multicode channel while in the other the coding is across both multicodes and time. In this paper, a performance/complexity analysis of decoding metrics and trellis structures for the two schemes is carried out. It is shown that the latter scheme can exploit the multicode diversity inherent in convolutionally coded direct sequence code division multiple access (DS-CDMA systems which employ minimum mean squared error (MMSE multiuser detectors. In particular, when the MMSE detector provides sufficiently different signal-to-interference ratios (SIRs for the multicode channels, coding across multicodes and time can obtain significant performance gain over coding across time, with nearly the same decoding complexity.

  5. Study of adaptive modulation and LDPC coding in multicarrier systems

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    An adaptive modulation (AM) algorithm is proposed and the application of the adapting algorithm together with low-density parity-check (LDPC) codes in multicarrier systems is investigated.The AM algorithm is based on minimizing the average bit error rate (BER) of systems,the combination of AM algorithm and LDPC codes with different code rates (half and three-fourths) are studied.The proposed AM algorithm with that of Fischer et al is compared.Simulation results show that the performance of the proposed AM algorithm is better than that of the Fischer's algorithm.The results also show that application of the proposed AM algorithm together with LDPC codes can greatly improve the performance of multicarrier systems.Results also show that the performance of the proposed algorithm is degraded with an increase in code rate when code length is the same.

  6. Implementation of the Resonance Analysis Code SAMMY

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The multi-level multi-channel R-matrix SAMMY code is used for making the resonance parameters,which was developed by Oak Ridge National Laboratory (ORNL), and widely used around the USA(ORELA, KAPL, LANL, TUNL...) and around the world (Belgium, Japan, France, Bulgaria, etc.).Thecode SAMMY is an important program to CNDC.

  7. TASS/SMR Code Topical Report for SMART Plant, Vol. I: Code Structure, System Models, and Solution Methods

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young Jong; Kim, Soo Hyoung; Kim, See Darl (and others)

    2008-10-15

    The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.

  8. Digital system detects binary code patterns containing errors

    Science.gov (United States)

    Muller, R. M.; Tharpe, H. M., Jr.

    1966-01-01

    System of square loop magnetic cores associated with code input registers to react to input code patterns by reference to a group of control cores in such a manner that errors are canceled and patterns containing errors are accepted for amplification and processing. This technique improves reception capabilities in PCM telemetry systems.

  9. COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

    International Nuclear Information System (INIS)

    COBRA-SFS is a general thermal-hydraulic analysis computer code for prediction of material temperatures and fluid conditions in a wide variety of systems. The code has been validated for analysis of spent fuel storage systems, as part of the Commercial Spent Fuel Management Program of the US Department of Energy. The code solves finite volume equations representing the conservation equations for mass, moment, and energy for an incompressible single-phase heat transfer fluid. The fluid solution is coupled to a finite volume solution of the conduction equation in the solid structure of the system. This document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part 1 describes the conservation equations, constitutive models, and solution methods used in the code. Part 2 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part 3 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available

  10. Criticality qualification of a new Monte Carlo code for reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Catsaros, N. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Gaveau, B. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Jaekel, M. [Laboratoire de Physique Theorique, Ecole Normale Superieure, 24 rue Lhomond, 75231 Paris (France); Maillard, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); CNRS-IDRIS, Bt 506, BP167, 91403 Orsay (France); CNRS-IN2P3, 3 rue Michel Ange, 75794 Paris (France); Maurel, G. [Faculte de Medecine, Universite Paris VI, 27 rue de Chaligny, 75012 Paris (France); MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Savva, P., E-mail: savvapan@ipta.demokritos.g [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Silva, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Varvayanni, M.; Zisis, Th. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece)

    2009-11-15

    In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

  11. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

    International Nuclear Information System (INIS)

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes

  12. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes.

  13. Comparison of Activation Analysis Codes between CINDER'90 and ORIGEN-S

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Dong; Choi, Hong Yeop; Lee, Yong Deok; Kim, Hodong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A Slowing Down Time Spectrometer (SDTS) system is the most feasible technology among the non-destructive techniques to directly analyze the content of isotopic fissile material. SDTS is necessary to a source neutron for inducing isotopic fissile fission. The source neutron is produced between the electron beam and a metal target by an (e,γ)(γ,n) reaction in the target. The target is required to have a high intensity neutron source through a proper target design. The status of activation on the designed target is analyzed through the activation code. Also, an activation evaluation of the material of the shielding facilities for SDTS system is required. The radioactivity intensity and kind of nuclides are measured through an activation analysis. ORIGEN-S and CINDER'90 codes are an activation code and are used in combination with the MCNPX code. ORIGEN-S code interprets a problem as one point about target. It cannot describe the geometry. CINDER'90 code can describe a 3D-geometry, and the result of CINDER'90 has high reliability when using a multi-group library. In this research, CINDER'90 was introduced as an activation analysis code and compared with the ORIGEN-S code. An activation analysis was conducted on the materials of the designed target. The ORIGEN-S and CINDER'90 code simulation results are provided for a selection of the activation analysis code. A Slowing Down Time Spectrometer (SDTS) system is a highly efficient technique in a nuclear material analysis. An activation analysis on the shielding and target material was required for the SDTS system. The activation results of CINDER'90 and ORIGEN-S codes were similar or different according to the nuclides because the cross section library of the codes is different. In utilizing the code, CINDER'90 code is more convenient than ORIGEN-S. It can describe the 3D-geometry, and therefore the activation information can be obtained by one simulation. The results of the activation

  14. REFLA-1D/MODE3: a computer code for reflood thermo-hydrodynamic analysis during PWR-LOCA

    International Nuclear Information System (INIS)

    This manual describes the REFLA-1D/MODE3 reflood system analysis code. This code can solve the core thermo-hydrodynamics under forced flooding conditions and gravity feed conditions in a system similar to FLECHT-SET Phase A. This manual describes the REFLA-1D/MODE3 models and provides application information required to utilize the code. (author)

  15. Path Weight Complementary Convolutional Code for Type-II Bit-Interleaved Coded Modulation Hybrid ARQ System

    Institute of Scientific and Technical Information of China (English)

    CHENG Yuxin; ZHANG Lei; YI Na; XIANG Haige

    2007-01-01

    Bit-interleaved coded modulation (BICM) is suitable to bandwidth-efficient communication systems. Hybrid automatic repeat request (HARQ) can provide more reliability to high-speed wireless data transmission. A new path weight complementary convolutional (PWCC) code used in the type-ll BICM-HARQ system is proposed. The PWCC code is composed of the original code and the complimentary code. The path in trellis with large hamming weight of the complimentary code is designed to compensate for the path in trellis with small hamming weight of the original code. Hence, both of the original code and the complimentary code can achieve the performance of the good code criterion of corresponding code rate. The throughput efficiency of the BICM-HARQ system wit PWCC code is higher than repeat code system, a little higher than puncture code system in low signal-to-noise ratio (SNR) values and much higher than puncture code system, the same as repeat code system in high SNR values. These results are confirmed by the simulation.

  16. Partial iterated function system-based fractal image coding

    Science.gov (United States)

    Wang, Zhou; Yu, Ying Lin

    1996-06-01

    A recent trend in computer graphics and image processing has been to use iterated function system (IFS) to generate and describe images. Barnsley et al. presented the conception of fractal image compression and Jacquin was the first to propose a fully automatic gray scale still image coding algorithm. This paper introduces a generalization of basic IFS, leading to a conception of partial iterated function system (PIFS). A PIFS operator is contractive under certain conditions and when it is applied to generate an image, only part of it is actually iteratedly applied. PIFS provides us a flexible way to combine fractal coding with other image coding techniques and many specific algorithms can be derived from it. On the basis of PIFS, we implement a partial fractal block coding (PFBC) algorithm and compare it with basic IFS based fractal block coding algorithm. Experimental results show that coding efficiency is improved and computation time is reduced while image fidelity does not degrade very much.

  17. Programme Code for Projecting of WDM Fiber Optic Sensor Systems

    OpenAIRE

    Probstner, R.; J. Turan

    1993-01-01

    Wavelength division multiplex (WDM) offers a potentially powerful technique for use within optical fibre sensor systems. The paper deals with short description of methodology and a programme code for WDM fiber optic sensor system projecting with use of CAD.

  18. Core-seis: a code for LMFBR core seismic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chellapandi, P.; Ravi, R.; Chetal, S.C.; Bhoje, S.B. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Reactor Group

    1995-12-31

    This paper deals with a computer code CORE-SEIS specially developed for seismic analysis of LMFBR core configurations. For demonstrating the prediction capability of the code, results are presented for one of the MONJU reactor core mock ups which deals with a cluster of 37 subassemblies kept in water. (author). 3 refs., 7 figs., 2 tabs.

  19. Improvement of QR Code Recognition Based on Pillbox Filter Analysis

    OpenAIRE

    Jia-Shing Sheu; Kai-Chung Teng

    2013-01-01

    The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the conte...

  20. Performance analysis of adaptive turbo coded modulation with time delay

    Institute of Scientific and Technical Information of China (English)

    伍守豪; 宋文涛; 罗汉文

    2004-01-01

    The method of data fitting is applied to obtain the BER expression for turbo coded modulation, and a fitting mathematical model is proposed, which resolves the problem that there is no exact BER expression for turbo coded modulation in performance analysis. With the time delay consideration, the performance of BER of adaptive turbo coded modulation is analyzed and simulated. The results show that adaptive turbo coded modulation is very sensitive to time delay. In order to meet the target BER requirement, the total time delay should be less than 0. 001/fD.

  1. System Measures Errors Between Time-Code Signals

    Science.gov (United States)

    Cree, David; Venkatesh, C. N.

    1993-01-01

    System measures timing errors between signals produced by three asynchronous time-code generators. Errors between 1-second clock pulses resolved to 2 microseconds. Basic principle of computation of timing errors as follows: central processing unit in microcontroller constantly monitors time data received from time-code generators for changes in 1-second time-code intervals. In response to any such change, microprocessor buffers count of 16-bit internal timer.

  2. Interactive computer code for dynamic and soil structure interaction analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mulliken, J.S.

    1995-12-01

    A new interactive computer code is presented in this paper for dynamic and soil-structure interaction (SSI) analyses. The computer program FETA (Finite Element Transient Analysis) is a self contained interactive graphics environment for IBM-PC`s that is used for the development of structural and soil models as well as post-processing dynamic analysis output. Full 3-D isometric views of the soil-structure system, animation of displacements, frequency and time domain responses at nodes, and response spectra are all graphically available simply by pointing and clicking with a mouse. FETA`s finite element solver performs 2-D and 3-D frequency and time domain soil-structure interaction analyses. The solver can be directly accessed from the graphical interface on a PC, or run on a number of other computer platforms.

  3. Development of environmental dose assessment system (EDAS) code of PC version

    CERN Document Server

    Taki, M; Kobayashi, H; Yamaguchi, T

    2003-01-01

    A computer code (EDAS) was developed to assess the public dose for the safety assessment to get the license of nuclear reactor operation. This code system is used for the safety analysis of public around the nuclear reactor in normal operation and severe accident. This code was revised and composed for personal computer user according to the Nuclear Safety Guidelines reflected the ICRP1990 recommendation. These guidelines are revised by Nuclear Safety Commission on March, 2001, which are 'Weather analysis guideline for the safety assessment of nuclear power reactor', 'Public dose around the facility assessment guideline corresponding to the objective value for nuclear power light water reactor' and 'Public dose assessment guideline for safety review of nuclear power light water reactor'. This code has been already opened for public user by JAERI, and English version code and user manual are also prepared. This English version code is helpful for international cooperation concerning the nuclear safety assessme...

  4. EquiFACS: The Equine Facial Action Coding System.

    Directory of Open Access Journals (Sweden)

    Jen Wathan

    Full Text Available Although previous studies of horses have investigated their facial expressions in specific contexts, e.g. pain, until now there has been no methodology available that documents all the possible facial movements of the horse and provides a way to record all potential facial configurations. This is essential for an objective description of horse facial expressions across a range of contexts that reflect different emotional states. Facial Action Coding Systems (FACS provide a systematic methodology of identifying and coding facial expressions on the basis of underlying facial musculature and muscle movement. FACS are anatomically based and document all possible facial movements rather than a configuration of movements associated with a particular situation. Consequently, FACS can be applied as a tool for a wide range of research questions. We developed FACS for the domestic horse (Equus caballus through anatomical investigation of the underlying musculature and subsequent analysis of naturally occurring behaviour captured on high quality video. Discrete facial movements were identified and described in terms of the underlying muscle contractions, in correspondence with previous FACS systems. The reliability of others to be able to learn this system (EquiFACS and consistently code behavioural sequences was high--and this included people with no previous experience of horses. A wide range of facial movements were identified, including many that are also seen in primates and other domestic animals (dogs and cats. EquiFACS provides a method that can now be used to document the facial movements associated with different social contexts and thus to address questions relevant to understanding social cognition and comparative psychology, as well as informing current veterinary and animal welfare practices.

  5. EquiFACS: The Equine Facial Action Coding System.

    Science.gov (United States)

    Wathan, Jen; Burrows, Anne M; Waller, Bridget M; McComb, Karen

    2015-01-01

    Although previous studies of horses have investigated their facial expressions in specific contexts, e.g. pain, until now there has been no methodology available that documents all the possible facial movements of the horse and provides a way to record all potential facial configurations. This is essential for an objective description of horse facial expressions across a range of contexts that reflect different emotional states. Facial Action Coding Systems (FACS) provide a systematic methodology of identifying and coding facial expressions on the basis of underlying facial musculature and muscle movement. FACS are anatomically based and document all possible facial movements rather than a configuration of movements associated with a particular situation. Consequently, FACS can be applied as a tool for a wide range of research questions. We developed FACS for the domestic horse (Equus caballus) through anatomical investigation of the underlying musculature and subsequent analysis of naturally occurring behaviour captured on high quality video. Discrete facial movements were identified and described in terms of the underlying muscle contractions, in correspondence with previous FACS systems. The reliability of others to be able to learn this system (EquiFACS) and consistently code behavioural sequences was high--and this included people with no previous experience of horses. A wide range of facial movements were identified, including many that are also seen in primates and other domestic animals (dogs and cats). EquiFACS provides a method that can now be used to document the facial movements associated with different social contexts and thus to address questions relevant to understanding social cognition and comparative psychology, as well as informing current veterinary and animal welfare practices.

  6. Arithmetic coding as a non-linear dynamical system

    Science.gov (United States)

    Nagaraj, Nithin; Vaidya, Prabhakar G.; Bhat, Kishor G.

    2009-04-01

    In order to perform source coding (data compression), we treat messages emitted by independent and identically distributed sources as imprecise measurements (symbolic sequence) of a chaotic, ergodic, Lebesgue measure preserving, non-linear dynamical system known as Generalized Luröth Series (GLS). GLS achieves Shannon's entropy bound and turns out to be a generalization of arithmetic coding, a popular source coding algorithm, used in international compression standards such as JPEG2000 and H.264. We further generalize GLS to piecewise non-linear maps (Skewed-nGLS). We motivate the use of Skewed-nGLS as a framework for joint source coding and encryption.

  7. Optical Code Processing System, Device, and its Application

    Directory of Open Access Journals (Sweden)

    Naoya Wada

    2010-02-01

    Full Text Available Recent progress of optical code processing technology_ is explained. Ultra-high speed time domain, spectral domain, hybrid_ domain, and multiple optical code processing deices and systems are shown. As application of these technologies, OCDMA-PON, OPS network, and ultra high-speed optical clock generation will be demonstrated.

  8. Global sensitivity analysis of the XUV-ABLATOR code

    Science.gov (United States)

    Nevrlý, Václav; Janku, Jaroslav; Dlabka, Jakub; Vašinek, Michal; Juha, Libor; Vyšín, Luděk.; Burian, Tomáš; Lančok, Ján.; Skřínský, Jan; Zelinger, Zdeněk.; Pira, Petr; Wild, Jan

    2013-05-01

    Availability of numerical model providing reliable estimation of the parameters of ablation processes induced by extreme ultraviolet laser pulses in the range of nanosecond and sub-picosecond timescales is highly desirable for recent experimental research as well as for practical purposes. Performance of the one-dimensional thermodynamic code (XUV-ABLATOR) in predicting the relationship of ablation rate and laser fluence is investigated for three reference materials: (i) silicon, (ii) fused silica and (iii) polymethyl methacrylate. The effect of pulse duration and different material properties on the model predictions is studied in the frame of this contribution for the conditions typical for two compact laser systems operating at 46.9 nm. Software implementation of the XUV-ABLATOR code including graphical user's interface and the set of tools for sensitivity analysis was developed. Global sensitivity analysis using high dimensional model representation in combination with quasi-random sampling was applied in order to identify the most critical input data as well as to explore the uncertainty range of model results.

  9. Concatenated coding systems employing a unit-memory convolutional code and a byte-oriented decoding algorithm

    Science.gov (United States)

    Lee, L.-N.

    1977-01-01

    Concatenated coding systems utilizing a convolutional code as the inner code and a Reed-Solomon code as the outer code are considered. In order to obtain very reliable communications over a very noisy channel with relatively modest coding complexity, it is proposed to concatenate a byte-oriented unit-memory convolutional code with an RS outer code whose symbol size is one byte. It is further proposed to utilize a real-time minimal-byte-error probability decoding algorithm, together with feedback from the outer decoder, in the decoder for the inner convolutional code. The performance of the proposed concatenated coding system is studied, and the improvement over conventional concatenated systems due to each additional feature is isolated.

  10. Improvement of QR Code Recognition Based on Pillbox Filter Analysis

    Directory of Open Access Journals (Sweden)

    Jia-Shing Sheu

    2013-04-01

    Full Text Available The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the contents in the QR code image. Many studies have used the pillbox filter (circular averaging filter method to simulate an out-of-focus image. This method is also used in this investigation to improve the recognition of a captured QR code image. A blurred QR code image is separated into nine levels. In the experiment, four different quantitative approaches are used to reconstruct and decode an out-of-focus QR code image. These nine reconstructed QR code images using methods are then compared. The final experimental results indicate improvements in identification.

  11. Ocean Thermal Energy Conversion power system development. Phase I: preliminary design. Final report. [ODSP-3 code; OTEC Steady-State Analysis Program

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-04

    The following appendices are included; Dynamic Simulation Program (ODSP-3); sample results of dynamic simulation; trip report - NH/sub 3/ safety precautions/accident records; trip report - US Coast Guard Headquarters; OTEC power system development, preliminary design test program report; medium turbine generator inspection point program; net energy analysis; bus bar cost of electricity; OTEC technical specifications; and engineer drawings. (WHK)

  12. MORSE Monte Carlo radiation transport code system

    International Nuclear Information System (INIS)

    For a number of years the MORSE user community has requested additional help in setting up problems using various options. The sample problems distributed with MORSE did not fully demonstrate the capability of the code. At Oak Ridge National Laboratory the code originators had a complete set of sample problems, but funds for documenting and distributing them were never available. Recently the number of requests for listings of input data and results for running some particular option the user was trying to implement has increased to the point where it is not feasible to handle them on an individual basis. Consequently it was decided to package a set of sample problems which illustrates more adequately how to run MORSE. This write-up may be added to Part III of the MORSE report. These sample problems include a combined neutron-gamma case, a neutron only case, a gamma only case, an adjoint case, a fission case, a time-dependent fission case, the collision density case, an XCHEKR run and a PICTUR run

  13. Outage probability analysis for superposition coded symmetric relaying

    Institute of Scientific and Technical Information of China (English)

    WU Yi; ZHENG Meng; FEI ZeSong; LARSSON Erik G.; KUANG JingMing

    2013-01-01

    Superposition coded symmetric relaying is a bandwidth-efficient cooperative scheme where each source node simultaneously transmits both its own "local" packet and "relay" packet that originated at its partner by adding the modulated local and relay signals in Euclidean space. This paper investigates the power allocation and outage probability of a superposition coded symmetric relaying system with finite-constellation signaling. We first derive the mutual information (MI) metrics for the system. The derived MI metrics consist of two parts: one represents the MI conveyed by the modulated signal corresponding to its own data, and the other represents the MI conveyed by the modulated signal corresponding to its partner's data. Using MI based effective signal-to-noise ratio mapping technique, we attain expressions for the outage probability. Furthermore, we discuss power allocation policies that minimize the outage probability. Simulation results are presented to verify the correctness of the outage probability analysis and the benefits of the power allocation.

  14. User's manual for seismic analysis code 'SONATINA-2V'

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, Satoshi; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

  15. User's guide for 10 CFR 61 impact analysis codes

    International Nuclear Information System (INIS)

    This document explains how to use the Impact Analysis Codes used in the Draft Environmental Impact Statement (DEIS) (NUREG-0782, Vol. 1-4) supporting 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste. The mathematical development of the impact Analysis Codes and other information necessary to understand the results of using the Codes is contained in the DEIS, and in a supporting document, Data Base for Radioactive Waste Management (NUREG/CR-1759, Vol. 1-3). This document was prepared with the intention of accompanying a computer magnetic tape containing the Impact Analysis Codes. A form is included at the end of this document which can be used to obtain such a tape

  16. CODING IN THE MAMMALIAN GUSTATORY SYSTEM

    Science.gov (United States)

    Carleton, Alan; Accolla, Riccardo; Simon, Sidney A.

    2010-01-01

    To understand gustatory physiology and associated dysfunctions it is important to know how stimuli placed in the mouth are encoded both in the periphery and in taste-related brain centres. The identification of distinct taste receptors, together with electrophysiological recordings and behavioural assessments in response to taste stimuli, suggest that information about distinct taste modalities (e.g., sweet versus bitter) are transmitted from the periphery to the brain via segregated pathways. In contrast, gustatory neurons throughout the brain are more broadly tuned, indicating that ensembles of neurons encode taste qualities. Recent evidence reviewed here suggests that the coding of gustatory stimuli is not immutable, but is dependant on a variety of factors including appetite regulating molecules and associative learning. PMID:20493563

  17. MULTIPLE TRELLIS CODED ORTHOGONAL TRANSMIT SCHEME FOR MULTIPLE ANTENNA SYSTEMS

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In this paper, a novel multiple trellis coded orthogonal transmit scheme is proposed to exploit transmit diversity in fading channels. In this scheme, a unique vector from a set of orthogonal vectors is assigned to each transmit antenna. Each of the output symbols from the multiple trellis encoder is multiplied with one of these orthogonal vectors and transmitted from corresponding transmit antennas. By correlating with corresponding orthogonal vectors, the receiver separates symbols transmitted from different transmit antennas.This scheme can be adopted in coherent/differential systems with any number of transmit antennas. It is shown that the proposed scheme encompasses the conventional trellis coded unitary space-time modulation based on the optimal cyclic group codes as a special case. We also propose two better designs over the conventional trellis coded unitary space-time modulation. The first design uses 8 Phase Shift Keying (8-PSK) constellations instead of 16 Phase Shift Keying (16-PSK) constellations in the conventional trellis coded unitary space-time modulation. As a result, the product distance of this new design is much larger than that of the conventional trellis coded unitary space-time modulation. The second design introduces constellations with multiple levels of amplitudes into the design of the multiple trellis coded orthogonal transmit scheme. For both designs, simulations show that multiple trellis coded orthogonal transmit schemes can achieve better performance than the conventional trellis coded unitary space-time schemes.

  18. Electronic manual of the nuclear characteristics analysis code-set for FBR

    International Nuclear Information System (INIS)

    Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

  19. Safety analysis and code development for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Development effort of computer codes applicable to nuclear fuel cycle facilities for assisting the task of NISA has been carried out. The work consists of 1) verification of criticality safety analysis codes : MVP and SCALE, 2) studies on burn-up credit applied methods, 3) preparation of non-uniformity effect calculation for criticality safety, 4) development of the new convenient library for shielding calculation based on JENDL-3.3 nuclear data, 5) development of a numerical simulation code DYMPL for analyzing abnormal transients of PUREX processes, 6) radiation dose evaluation code development for reprocessing facilities, 7) updating the dose evaluation data for the probabilistic environmental assessment code MACCS2-JF by emergency scenario. (author)

  20. Benchmarking Of Improved DPAC Transient Deflagration Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, James E.; Hensel, Steve J.

    2013-03-21

    The transient deflagration code DPAC (Deflagration Pressure Analysis Code) has been upgraded for use in modeling hydrogen deflagration transients. The upgraded code is benchmarked using data from vented hydrogen deflagration tests conducted at the HYDRO-SC Test Facility at the University of Pisa. DPAC originally was written to calculate peak deflagration pressures for deflagrations in radioactive waste storage tanks and process facilities at the Savannah River Site. Upgrades include the addition of a laminar flame speed correlation for hydrogen deflagrations and a mechanistic model for turbulent flame propagation, incorporation of inertial effects during venting, and inclusion of the effect of water vapor condensation on vessel walls. In addition, DPAC has been coupled with CEA, a NASA combustion chemistry code. The deflagration tests are modeled as end-to-end deflagrations. The improved DPAC code successfully predicts both the peak pressures during the deflagration tests and the times at which the pressure peaks.

  1. Thermal-hydraulic Analysis and Code Assessment for Reactor Vessel Upper-head Small Break LOCA using SPACE code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper-head as a result of circumferential cracking of a Control Red Drive Mechanism (CRDM) penetration nozzle. Several experimental tests have been performed at the large scale test facility to simulate the behavior of a PWR during an upper-head SBLOCA. Organization for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1 was performed with a break size equivalent to 1.9% cold leg break. Additionally, analysis of an upper-head SBLOCA with high pressure safety injection failed in a Westinghouse PWR was examined taking into account different accident management actions and conditions in order to check the suitability. In this study, the thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 (Optimized Power Reactor 1000 MWe) using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. Inspections of existing nuclear power plants have pointed out the possibility of small break loss of coolant accidents (SBLOCAs) were initiated by a small break located in the upper-head of the reactor pressure vessel. The thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 plant using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. The prediction showed good agreement with the MARS

  2. Thermal-hydraulic Analysis and Code Assessment for Reactor Vessel Upper-head Small Break LOCA using SPACE code

    International Nuclear Information System (INIS)

    In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper-head as a result of circumferential cracking of a Control Red Drive Mechanism (CRDM) penetration nozzle. Several experimental tests have been performed at the large scale test facility to simulate the behavior of a PWR during an upper-head SBLOCA. Organization for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1 was performed with a break size equivalent to 1.9% cold leg break. Additionally, analysis of an upper-head SBLOCA with high pressure safety injection failed in a Westinghouse PWR was examined taking into account different accident management actions and conditions in order to check the suitability. In this study, the thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 (Optimized Power Reactor 1000 MWe) using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. Inspections of existing nuclear power plants have pointed out the possibility of small break loss of coolant accidents (SBLOCAs) were initiated by a small break located in the upper-head of the reactor pressure vessel. The thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 plant using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. The prediction showed good agreement with the MARS

  3. Codes, standards, and PV power systems. A 1996 status report

    Energy Technology Data Exchange (ETDEWEB)

    Wiles, J

    1996-06-01

    As photovoltaic (PV) electrical power systems gain increasing acceptance for both off-grid and utility-interactive applications, the safety, durability, and performance of these systems gains in importance. Local and state jurisdictions in many areas of the country require that all electrical power systems be installed in compliance with the requirements of the National Electrical Code{reg_sign} (NEC{reg_sign}). Utilities and governmental agencies are now requiring that PV installations and components also meet a number of Institute of Electrical and Electronic Engineers (IEEE) standards. PV installers are working more closely with licensed electricians and electrical contractors who are familiar with existing local codes and installation practices. PV manufacturers, utilities, balance of systems manufacturers, and standards representatives have come together to address safety and code related issues for future PV installations. This paper addresses why compliance with the accepted codes and standards is needed and how it is being achieved.

  4. ARC Code TI: Optimal Alarm System Design and Implementation

    Data.gov (United States)

    National Aeronautics and Space Administration — An optimal alarm system can robustly predict a level-crossing event that is specified over a fixed prediction horizon. The code contained in this packages provides...

  5. Study on New Concatenated Code in WDM Optical Transmission Systems

    Institute of Scientific and Technical Information of China (English)

    YUAN Jian-guo; JIANG Ze; MAO You-ju; YE Wen-wei

    2007-01-01

    A new concatenated code of RS(255,239)+BCH(2 040,1 930) code to be suitable for WDM optical transmission systems is proposed.The simulation results show that this new concatenated code,compared with the RS(255,239)+CSOC(k0/n0=6/7,J=8) code in ITU-T G.75.1,has a lower redundancy and better error-correction performance,furthermore,its net coding gain(NCG) is respectively 0.46 dB,0.43 dB more than that of RS(255,239)+CSOC(k0/n0 =6/7,J=8) code and BCH(3 860,3 824)+BCH(2 040,1 930) code in ITU-T G.75.1 at the third iteration for the bit error rate(BER) of 10-12.Therefore,the new super forward error correction(Super-FEC) concatenated code can be better used in ultra long-haul,ultra large-capacity and ultra high-speed WDM optical communication systems.

  6. Code-modulated interferometric imaging system using phased arrays

    Science.gov (United States)

    Chauhan, Vikas; Greene, Kevin; Floyd, Brian

    2016-05-01

    Millimeter-wave (mm-wave) imaging provides compelling capabilities for security screening, navigation, and bio- medical applications. Traditional scanned or focal-plane mm-wave imagers are bulky and costly. In contrast, phased-array hardware developed for mass-market wireless communications and automotive radar promise to be extremely low cost. In this work, we present techniques which can allow low-cost phased-array receivers to be reconfigured or re-purposed as interferometric imagers, removing the need for custom hardware and thereby reducing cost. Since traditional phased arrays power combine incoming signals prior to digitization, orthogonal code-modulation is applied to each incoming signal using phase shifters within each front-end and two-bit codes. These code-modulated signals can then be combined and processed coherently through a shared hardware path. Once digitized, visibility functions can be recovered through squaring and code-demultiplexing operations. Pro- vided that codes are selected such that the product of two orthogonal codes is a third unique and orthogonal code, it is possible to demultiplex complex visibility functions directly. As such, the proposed system modulates incoming signals but demodulates desired correlations. In this work, we present the operation of the system, a validation of its operation using behavioral models of a traditional phased array, and a benchmarking of the code-modulated interferometer against traditional interferometer and focal-plane arrays.

  7. The Marriage of Residential Energy Codes and Rating Systems: Conflict Resolution or Just Conflict?

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Zachary T.; Mendon, Vrushali V.

    2014-08-21

    After three decades of coexistence at a distance, model residential energy codes and residential energy rating systems have come together in the 2015 International Energy Conservation Code. At the October, 2013, International Code Council’s Public Comment Hearing, a new compliance path based on an Energy Rating Index was added to the IECC. Although not specifically named in the code, RESNET’s HERS rating system is the likely candidate Index for most jurisdictions. While HERS has been a mainstay in various beyond-code programs for many years, its direct incorporation into the most popular model energy code raises questions about the equivalence of a HERS-based compliance path and the traditional IECC performance compliance path, especially because the two approaches use different efficiency metrics, are governed by different simulation rules, and have different scopes with regard to energy impacting house features. A detailed simulation analysis of more than 15,000 house configurations reveals a very large range of HERS Index values that achieve equivalence with the IECC’s performance path. This paper summarizes the results of that analysis and evaluates those results against the specific Energy Rating Index values required by the 2015 IECC. Based on the home characteristics most likely to result in disparities between HERS-based compliance and performance path compliance, potential impacts on the compliance process, state and local adoption of the new code, energy efficiency in the next generation of homes subject to this new code, and future evolution of model code formats are discussed.

  8. Recording and Replaying System Specific, Source Code Transformations

    OpenAIRE

    Santos, Gustavo; Etien, Anne; Anquetil, Nicolas; Ducasse, Stéphane; Tulio Valente, Marco

    2015-01-01

    International audience During its lifetime, a software system is under continuous maintenance to remain useful. Maintenance can be achieved in activities such as adding new features, fixing bugs, improving the system's structure, or adapting to new APIs. In such cases, developers sometimes perform sequences of code changes in a systematic way. These sequences consist of small code changes (e.g., create a class, then extract a method to this class), which are applied to groups of related co...

  9. Optimal Coding Predicts Attentional Modulation of Activity in Neural Systems

    OpenAIRE

    Jaramillo, Santiago; Pearlmutter, Barak A.

    2007-01-01

    Neuronal activity in response to a fixed stimulus has been shown to change as a function of attentional state, implying that the neural code also changes with attention. We propose an information-theoretic account of such modulation: that the nervous system adapts to optimally encode sensory stimuli while taking into account the changing relevance of different features. We show using computer simulation that such modulation emerges in a coding system informed about the uneven relevance of ...

  10. Guide to Using Onionskin Analysis Code (U)

    Energy Technology Data Exchange (ETDEWEB)

    Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group; Morzinski, Jerome Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group

    2016-09-15

    This document is a guide to using R-code written for the purpose of analyzing onionskin experiments. We expect the user to be very familiar with statistical methods and the R programming language. For more details about onionskin experiments and the statistical methods mentioned in this document see Storlie, Fugate, et al. (2013). Engineers at LANL experiment with detonators and high explosives to assess performance. The experimental unit, called an onionskin, is a hemisphere consisting of a detonator and a booster pellet surrounded by explosive material. When the detonator explodes, a streak camera mounted above the pole of the hemisphere records when the shock wave arrives at the surface. The output from the camera is a two-dimensional image that is transformed into a curve that shows the arrival time as a function of polar angle. The statistical challenge is to characterize a baseline population of arrival time curves and to compare the baseline curves to curves from a new, so-called, test series. The hope is that the new test series of curves is statistically similar to the baseline population.

  11. New York State Code Adoption Analysis: Lighting Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Richman, Eric E.

    2004-10-20

    The adoption of the IECC 2003 Energy code will include a set of Lighting Power Density (LPD) values that are effectively a subset of the values in Addendum g to the ASHRAE/IESNA/ANSI 90.1-2001 Standard which will soon be printed as part of the 90.1-2004 version. An analysis of the effectiveness of this adoption for New York State can be provided by a direct comparison of these values with existing LPD levels represented in the current IECC 2000 code, which are themselves a subset of the current ASHRAE/IESNA/ANSI 90.1-2001 Standard (without addenda). Because the complete ASHRAE 2001 and 2004 sets of LPDs are supported by a set of detailed models, they are best suited to provide the basis for an analysis comparison of the two code levels of lighting power density stringency. It is important to note that this kind of analysis is a point-to-point comparison where a fixed level of real world activity is assumed. It is understood that buildings are not built precisely to code levels and that actual percentage of compliance above and below codes will vary among individual buildings and building types. However, without specific knowledge of this real world activity for all buildings in existence and in the future (post-code adoption) it is not possible to analyze actual effects of code adoption. However, it is possible to compare code levels and determine the potential effect of changes from one code requirement level to another. This is the comparison and effectiveness assessment

  12. Interface design of VSOP'94 computer code for safety analysis

    International Nuclear Information System (INIS)

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects

  13. Interface design of VSOP'94 computer code for safety analysis

    Science.gov (United States)

    Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

    2014-09-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

  14. Software Security Analysis : Managing source code audit

    OpenAIRE

    Persson, Daniel; Baca, Dejan

    2004-01-01

    Software users have become more conscious of security. More people have access to Internet and huge databases of security exploits. To make secure products, software developers must acknowledge this threat and take action. A first step is to perform a software security analysis. The software security analysis was performed using automatic auditing tools. An experimental environment was constructed to check if the findings were exploitable or not. Open source projects were used as reference to...

  15. Digital Image Analysis for Detechip Code Determination

    OpenAIRE

    Marcus Lyon; Wilson, Mark V.; Kerry A. Rouhier; David J. Symonsbergen; Kiran Bastola; Ishwor Thapa; Holmes, Andrea E.; Sharmin M. Sikich; Abby Jackson

    2012-01-01

    DETECHIP® is a molecular sensing array used for identification of a large variety of substances. Previous methodology for the analysis of DETECHIP® used human vision to distinguish color changes induced by the presence of the analyte of interest. This paper describes several analysis techniques using digital images of DETECHIP® . Both a digital camera and flatbed desktop photo scanner were used to obtain Jpeg images. Color information within these digital images was obt...

  16. Digital Image Analysis for Detechip Code Determination

    OpenAIRE

    Marcus Lyon; Wilson, Mark V.; Holmes, Andrea E.; Sharmin M. Sikich; Abby Jackson; Kerry A. Rouhier; David J. Symonsbergen; Kiran Bastola

    2012-01-01

    DETECHIP® is a molecular sensing array used for identification of a large variety of substances. Previous methodology for the analysis of DETECHIP® used human vision to distinguish color changes induced by the presence of the analyte of interest. This paper describes several analysis techniques using digital images of DETECHIP®. Both a digital camera and flatbed desktop photo scanner were used to obtain Jpeg images. Color information within these digital images was obtained through the me...

  17. Digital Image Analysis for DETCHIP® Code Determination

    OpenAIRE

    Lyon, Marcus; Wilson, Mark V.; Kerry A. Rouhier; David J. Symonsbergen; Bastola, Kiran; Thapa, Ishwor; Holmes, Andrea E.; Sharmin M. Sikich; Jackson, Abby

    2012-01-01

    DETECHIP® is a molecular sensing array used for identification of a large variety of substances. Previous methodology for the analysis of DETECHIP® used human vision to distinguish color changes induced by the presence of the analyte of interest. This paper describes several analysis techniques using digital images of DETECHIP®. Both a digital camera and flatbed desktop photo scanner were used to obtain Jpeg images. Color information within these digital images was obtained through the measur...

  18. A static analysis tool set for assembler code verification

    International Nuclear Information System (INIS)

    Software Verification and Validation (V and V) is an important step in assuring reliability and quality of the software. The verification of program source code forms an important part of the overall V and V activity. The static analysis tools described here are useful in verification of assembler code. The tool set consists of static analysers for Intel 8086 and Motorola 68000 assembly language programs. The analysers examine the program source code and generate information about control flow within the program modules, unreachable code, well-formation of modules, call dependency between modules etc. The analysis of loops detects unstructured loops and syntactically infinite loops. Software metrics relating to size and structural complexity are also computed. This report describes the salient features of the design, implementation and the user interface of the tool set. The outputs generated by the analyser are explained using examples taken from some projects analysed by this tool set. (author). 7 refs., 17 figs

  19. Safety analysis and the code development on radioactive waste disposal

    International Nuclear Information System (INIS)

    In order to confirm the long-term safety concerning sub-surface disposal, we studied the function about the climatic and topographic changes included in three-dimensional groundwater flow analysis code 3D-SEEP. And, we studied the methods of the groundwater flow analysis and particle tracking analysis in consideration of long-term phenomenon. Moreover, we made the trial calculations of the long-term transient analysis using this function. As a result, we found the adaptation range of the code and the differences from the results obtained by the steady state analysis. As a reflection of new knowledge about the particle tracking analysis, we carried out the trial calculation which was adapted in the analysis technique in consideration of the geometry model which changes with time progress. As a result, we found the differences from the results obtained by the conventional method, and the present subjects. We introduced and improved the groundwater flow and nuclide migration analysis code MH-FLOW using the mixed hybrid finite element method. This analysis code was developed for the purpose of obtaining a solution with sufficient accuracy even for the heterogeneous place where coefficients of permeability is greatly different. Moreover, we used MH-FLOW for the benchmark problem defined in an international project, and compared results with those obtained by the project. As a result, we checked the validity of MH-FLOW. (author)

  20. Miniature neutron source reactor burnup calculations using IRBURN code system

    International Nuclear Information System (INIS)

    Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.

  1. Comparative study of Barcode, QR-code and RFID System

    Directory of Open Access Journals (Sweden)

    Trupti Lotlikar

    2013-09-01

    Full Text Available Wireless sensors are standard measurement tools equipped with transmitters to convert signals from process control instruments into a radio transmission. The radio signal is interpreted by a receiver which then converts the wireless signal to a specific, desired output, such as an analog current or data analysis via computer software. The paper gives a brief on wireless sensors and their types like Barcode, QR code, RFID along with their characteristics and working components. The Barcode is an optical machine-readable representation of data relating to the object to which it is attached. On the other hand the Radio-frequency identification (RFID is the use of a wireless non-contact system that uses radio-frequency electromagnetic fields to transfer data from a tag attached to an object, for the purposes of automatic identification and tracking. Quick response (QR codes are a very convenient way to display a small bit of information that is easily scanned and processed typically by mobile devices allowing physical items to almost become interactive, by providing information that is easily scanned like a website URL. Finally this paper will compare all the three technologies on various grounds like durability, cost, information capacity, read range etc. to determine best out of it.

  2. JEMs and incompatible occupational coding systems: Effect of manual and automatic recoding of job codes on exposure assignment

    NARCIS (Netherlands)

    Koeman, T.; Offermans, N.S.M.; Christopher-De Vries, Y.; Slottje, P.; Brandt, P.A. van den; Goldbohm, R.A.; Kromhout, H.; Vermeulen, R.

    2013-01-01

    Background: In epidemiological studies, occupational exposure estimates are often assigned through linkage of job histories to job-exposure matrices (JEMs). However, available JEMs may have a coding system incompatible with the coding system used to code the job histories, necessitating a translatio

  3. Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

  4. Unidirectional Error Correcting Codes for Memory Systems: A Comparative Study

    CERN Document Server

    Al-Ani, Muzhir

    2010-01-01

    In order to achieve fault tolerance, highly reliable system often require the ability to detect errors as soon as they occur and prevent the speared of erroneous information throughout the system. Thus, the need for codes capable of detecting and correcting byte errors are extremely important since many memory systems use b-bit-per-chip organization. Redundancy on the chip must be put to make fault-tolerant design available. This paper examined several methods of computer memory systems, and then a proposed technique is designed to choose a suitable method depending on the organization of memory systems. The constructed codes require a minimum number of check bits with respect to codes used previously, then it is optimized to fit the organization of memory systems according to the requirements for data and byte lengths.

  5. Monte Carlo Code System Development for Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Shim, Hyung Jin; Han, Beom Seok; Park, Ho Jin; Park, Dong Gyu [Seoul National University, Seoul (Korea, Republic of)

    2007-03-15

    We have implemented the composition cell class and the use cell to MCCARD for hierarchy input processing. For the inputs of KALlMER-600 core consisted of 336 assemblies, we require the geometric data of 91,056 pin cells. Using hierarchy input processing, it was observed that the system geometries are correctly handled with the geometric data of total 611 cells; 2 cells for fuel rods, 2 cells for guide holes, 271 translation cells for rods, and 336 translation cells for assemblies. We have developed monte carlo decay-chain models based on decay chain model of REBUS code for liquid metal reactor analysis. Using developed decay-chain models, the depletion analysis calculations have performed for the homogeneous and heterogeneous model of KALlMER-600. The k-effective for the depletion analysis agrees well with that of REBUS code. and the developed decay chain models shows more efficient performance for time and memories, as compared with the existing decay chain model The chi-square criterion has been developed to diagnose the temperature convergence for the MC TjH feedback calculations. From the application results to the KALlMER pin and fuel assembly problem, it is observed that the new criterion works well Wc have applied the high efficiency variance reduction technique by splitting Russian roulette to estimate the PPPF of the KALIMER core at BOC. The PPPF of KALlMER core at BOC is 1.235({+-}0.008). The developed technique shows four time faster calculation, as compared with the existin2 calculation Subject Keywords Monte Carlo

  6. Two-dimensional disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)

  7. Construction Zero Cross Correlation Code using Permutation Matrix for SAC-OCDMA Systems

    OpenAIRE

    Nisar, K. S.

    2016-01-01

    This paper present a new method for constructing zero cross correlation code with the help of permutation matrices. The benefits of this newly proposed code are easy way code construction, the code weight exist for every natural number and the code length is acceptable. The numerical comparison shows that the proposed code has better or compatible code length compared with other existing zero cross correlation code in Optical Spectrum Code Division Multiple Access (OSCDMA) systems.

  8. Automatic code generation for distributed robotic systems

    International Nuclear Information System (INIS)

    Hetero Helix is a software environment which supports relatively large robotic system development projects. The environment supports a heterogeneous set of message-passing LAN-connected common-bus multiprocessors, but the programming model seen by software developers is a simple shared memory. The conceptual simplicity of shared memory makes it an extremely attractive programming model, especially in large projects where coordinating a large number of people can itself become a significant source of complexity. We present results from three system development efforts conducted at Oak Ridge National Laboratory over the past several years. Each of these efforts used automatic software generation to create 10 to 20 percent of the system

  9. Hydrogen detection systems leak response codes

    International Nuclear Information System (INIS)

    A loss in tightness of a water tube inside a Steam Generator Unit of a Fast Reactor is usually monitored by hydrogen detection systems. Such systems have demonstrated in the past their ability to detect a leak in a SGU. However, the increase in size of the SGU or the choice of ferritic material entails improvement of these systems in order to avoid secondary leak or to limit damages to the tube bundle. The R and D undertaken in France on this subject is presented. (author). 11 refs, 10 figs

  10. Building Secure Networked Systems with Code Attestation

    Science.gov (United States)

    Perrig, Adrian

    Attestation is a promising approach for building secure systems. The recent development of a Trusted Platform Module (TPM) by the Trusted Computing Group (TCG) that is starting to be deployed in common laptop and desktop platforms is fueling research in attestation mechanisms. In this talk, we will present approaches on how to build secure systems with advanced TPM architectures. In particular, we have designed an approach for fine-grained attestation that enables the design of efficient secure distributed systems, and other network protocols.We demonstrate this approach by designing a secure routing protocol.

  11. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    International Nuclear Information System (INIS)

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  12. JPEG2000 COMPRESSION CODING USING HUMAN VISUAL SYSTEM MODEL

    Institute of Scientific and Technical Information of China (English)

    Xiao Jiang; Wu Chengke

    2005-01-01

    In order to apply the Human Visual System (HVS) model to JPEG2000 standard,several implementation alternatives are discussed and a new scheme of visual optimization isintroduced with modifying the slope of rate-distortion. The novelty is that the method of visual weighting is not lifting the coefficients in wavelet domain, but is complemented by code stream organization. It remains all the features of Embedded Block Coding with Optimized Truncation (EBCOT) such as resolution progressive, good robust for error bit spread and compatibility of lossless compression. Well performed than other methods, it keeps the shortest standard codestream and decompression time and owns the ability of VIsual Progressive (VIP) coding.

  13. Report on nuclear industry quality assurance procedures for safety analysis computer code development and use

    International Nuclear Information System (INIS)

    As a result of a request from Commissioner V. Gilinsky to investigate in detail the causes of an error discovered in a vendor Emergency Core Cooling System (ECCS) computer code in March, 1978, the staff undertook an extensive investigation of the vendor quality assurance practices applied to safety analysis computer code development and use. This investigation included inspections of code development and use practices of the four major Light Water Reactor Nuclear Steam Supply System vendors and a major reload fuel supplier. The conclusion reached by the staff as a result of the investigation is that vendor practices for code development and use are basically sound. A number of areas were identified, however, where improvements to existing vendor procedures should be made. In addition, the investigation also addressed the quality assurance (QA) review and inspection process for computer codes and identified areas for improvement

  14. Disparities in obesity rates: Analysis by ZIP code area

    OpenAIRE

    Drewnowski, Adam; Rehm, Colin D.; Solet, David

    2007-01-01

    Obesity in the USA has been linked to individual income and education. Less is known about its geographic distribution. The goal of this study was to determine whether obesity rates in King County, Seattle, Washington state, at the ZIP code scale were associated with area-based measures of socioeconomic status and wealth. Data from the Behavioral Risk Factor Surveillance System were analyzed. At the ZIP code scale, crude obesity rates varied six-fold. In a model adjusting for covariates and s...

  15. Effects of bar coding on a pharmacy stock replenishment system.

    Science.gov (United States)

    Chester, M I; Zilz, D A

    1989-07-01

    A bar-code stock ordering system installed in the ambulatory-care pharmacy and sterile products area of a hospital pharmacy was compared with a manual paper system to quantify overall time demands and determine the error rate associated with each system. The bar-code system was implemented in the ambulatory-care pharmacy in November 1987 and in the sterile products area in January 1988. It consists of a Trakker 9440 transaction manager with a digital scanner; labels are printed with a dot matrix printer. Electronic scanning of bar-code labels and entry of the amount required using the key-pad on the transaction manager replaced use of a preprinted form for ordering items. With the bar-code system, ordering information is transferred electronically via cable to the pharmacy inventory computer; with the manual system, this information was input by a stockroom technician. To compare the systems, the work of technicians in the ambulatory-care pharmacy and sterile products area was evaluated before and after implementation of the bar-code system. The time requirements for information gathering and data transfer were recorded by direct observation; the prevalence of errors under each system was determined by comparing unprocessed ordering information with the corresponding computer-generated "pick lists" (itemized lists including the amount of each product ordered). Time consumed in extra trips to the stockroom to replace out-of-stock items was self-reported. Significantly less time was required to order stock and transfer data to the pharmacy inventory computer with the bar-code system than with the manual system.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:2757044

  16. Density and water content corrections in the gamma count rate of a PGNAA system for cement raw material analysis using the MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Carlos; Salgado, Jose [Nuclear and Technological Institute, Sacavem (Portugal); Leitao, Francisco [Technical Centre, Cimpor, Lisbon (Portugal)

    1998-05-11

    A MCNP simulation study for a prompt gamma neutron activation analysis system for on-line characterisation of cement raw materials has been carried out. A neutron source is located below a conveyor belt. Two detector banks were used: a lower bank positioned symmetrically around the source to detect {gamma}-rays emitted downwards; an upper bank detects the radiation emitted upwards. The count rate of both detector banks for a given composition depends on the bulk density and water content. This paper reports a few corrections which linearise the dependence of the corrected count rate on the mass content.

  17. Development of computer code packages for molten salt reactor core analysis

    International Nuclear Information System (INIS)

    This paper presents the implementations of the Oak Ridge National Laboratory (ORNL) approach for Molten Salt Reactor (MSR) core analysis with two nuclear reactor core analysis computer code systems. The first code system has been set up with the MCNP6 Monte Carlo code, its depletion module CINDER90 and the PYTHON script language. The second code system has been set up with the NEWT transport calculation module and ORIGEN depletion module connected by TRITON sequence in SCALE code, and the PYTHON script language. The PYTHON script language is used for implementing the online reprocessing of molten-salt fuel, and feeding new fertile material in the computer code simulations. In this paper, simplified nuclear reactor core models of a Molten Salt Breeder Reactor (MSBR), designed by ORNL in the 1960's, and FUJI-U3 designed by Toyohashi University of Technology (TUT) in the 2000's, were analyzed by the two code systems. Using these, various reactor design parameters of the MSRs were compared, such as the multiplication factor, breeding ratio, amount of material, total feeding, neutron flux distribution, and temperature coefficient. (author)

  18. 利用自适应编码调制的无线网络控制系统的分析和设计%Analysis and Design of Wireless Networked Control System Utilizing Adaptive Coded Modulation

    Institute of Scientific and Technical Information of China (English)

    杨丽; 郑新平; 龙承念; 罗小元

    2009-01-01

    In this paper, we explore the analysis and design of wireless networked control system (WNCS) utilizing adaptive coded modulation (ACM) schemes, which can improve the energy efficiency and increase data rate over a fading channel. To capture the characteristics of varying rate, interference, and routing in wireless transmission channel, we introduce the concept of equivalent delay (ED). Based on the time-varying network condition, the analytic lower and upper bounds of EDs are given. Whereafter, WNCS is modelled as a discrete-time system with time-varying input delay. Sufficient stabilization condition of the closed-loop WNCS is derived by making use of novel techniques of time-delay system. Finally, the numerical result shows the validity of our proposed control strategies.

  19. Maximum Likelihood Blind Channel Estimation for Space-Time Coding Systems

    Directory of Open Access Journals (Sweden)

    Hakan A. Çırpan

    2002-05-01

    Full Text Available Sophisticated signal processing techniques have to be developed for capacity enhancement of future wireless communication systems. In recent years, space-time coding is proposed to provide significant capacity gains over the traditional communication systems in fading wireless channels. Space-time codes are obtained by combining channel coding, modulation, transmit diversity, and optional receive diversity in order to provide diversity at the receiver and coding gain without sacrificing the bandwidth. In this paper, we consider the problem of blind estimation of space-time coded signals along with the channel parameters. Both conditional and unconditional maximum likelihood approaches are developed and iterative solutions are proposed. The conditional maximum likelihood algorithm is based on iterative least squares with projection whereas the unconditional maximum likelihood approach is developed by means of finite state Markov process modelling. The performance analysis issues of the proposed methods are studied. Finally, some simulation results are presented.

  20. Error correcting codes for binary unitary channels on multipartite quantum systems

    CERN Document Server

    Choi, M D; Kribs, D W; Zyczkowski, K; Choi, Man-Duen; Holbrook, John A.; Kribs, David W.; Zyczkowski, Karol

    2006-01-01

    We conduct an analysis of ideal error correcting codes for randomized unitary channels determined by two unitary error operators -- what we call ``binary unitary channels'' -- on multipartite quantum systems. In a wide variety of cases we give a complete description of the code structure for such channels. Specifically, we find a practical geometric technique to determine the existence of codes of arbitrary dimension, and then derive an explicit construction of codes of a given dimension when they exist. For instance, given any binary unitary noise model on an n-qubit system, we design codes that support n-2 qubits. We accomplish this by verifying a conjecture for higher rank numerical ranges of normal operators in many cases.

  1. The FORTRAN static source code analyzer program (SAP) system description

    Science.gov (United States)

    Decker, W.; Taylor, W.; Merwarth, P.; Oneill, M.; Goorevich, C.; Waligora, S.

    1982-01-01

    A source code analyzer program (SAP) designed to assist personnel in conducting studies of FORTRAN programs is described. The SAP scans FORTRAN source code and produces reports that present statistics and measures of statements and structures that make up a module. The processing performed by SAP and of the routines, COMMON blocks, and files used by SAP are described. The system generation procedure for SAP is also presented.

  2. Impact testing and analysis for structural code benchmarking

    International Nuclear Information System (INIS)

    Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks, R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem

  3. Performance Evaluation of Hybrid ARQ with Code Combining in Packet-Oriented CDMA System

    Institute of Scientific and Technical Information of China (English)

    CHENQingchun; FANPingzhi

    2004-01-01

    In this paper, an extended SNR (signal to noise ratio) concept is proposed to explicate the contribution of code combining to the performance improvement of hybrid ARQ (Automatic repeat request) over the additive white Gaussian noise channel. By extending the Pursley's SNR analysis to hybrid ARQ with code combining in packet-oriented CDMA (Code division multiple access)system, the extended SNR formula is derived, which describes explicitly the SNR variation of the code symbol involved in code combining. It is revealed that the extended SNR formula includes Pursley's SNR formula as a specialcase. Moreover, it is shown that the effective SNR of the combined symbol is increased by a coefficient, which is proportional to the number of repeated replicas involved in the code combining. Based on the extended SNR formula and the resultant SNR variation, a quasi-analytical approximation method is proposed for the performance evaluation of hybrid ARQ with code combining. The residual error rates, average transmission number together with throughput performance are presented by means of numerical analysis and through simulations. It is validated that the extended SNR formula and the resultant quasi-analytical approximations offer a simplified routine to estimate the performance of hybrid ARQ with code combining, particularly for the applications whose reliability performance with respect to the FEC counterpart system could be numerically calculated or evaluated through simulations.

  4. PCLOOK: an interactive code for spectral analysis

    International Nuclear Information System (INIS)

    The present work describes an interactive programme for the analysis of spectra developed to run in a PC platform. PCLOOK has a graphic interface that allows the user to get access to different functions using the mouse or directly typing commands. In this way one can switch to a suitable required environment to manage the histograms reassembling in this way a spectrum calculator.The PCLOOK programme was mainly developed to use in nuclear physics applications, but it is also possible to modify it with relative little effort to adapt it to other applications. It was written in Microsoft's BASIC 7.1 installed in a 33MHz 486 Everex PC. For a proper operation an ordinary VGA display and mouse are needed. The memory requirements depend on the size and number of the user defined spectra; for instance, for twenty 2048 channels spectra the available memory space must be 320 KBytes. (author). 5 figs

  5. EVAPRED - A CODE FOR FATIGUE ANALYSIS OPTIMIZATION

    Directory of Open Access Journals (Sweden)

    Dorin LOZICI-BRÎNZEI

    2010-03-01

    Full Text Available The fatigue can be, in fact, defined as: “failure under a repeated or otherwise varying load, which never reaches a level sufficient to cause failure in a single application”.Physical testing is clearly unrealistic for every design component. In most applications, fatigue-safe life design requires the prediction of the component fatigue life that accounts for predicted service loads and materials. The primary tool for both understanding and being able to predict and avoid fatigue has proven to be the finite element analysis (FEA. Computer-aided engineering (CAE programs use three major methods to determine the total fatigue life: Stress life (SN, Strain life (EN and Fracture Mechanics (FM. FEA can predict stress concentration areas and can help design engineers to predict how long their designs are likely to last before experiencing the onset of fatigue.

  6. QR Codes in the Library: Are They Worth the Effort? Analysis of a QR Code Pilot Project

    OpenAIRE

    Andrew M. Wilson

    2012-01-01

    The literature is filled with potential uses for Quick Response (QR) codes in the library. Setting, but few library QR code projects have publicized usage statistics. A pilot project carried out in the Eda Kuhn Loeb Music Library of the Harvard College Library sought to determine whether library patrons actually understand and use QR codes. Results and analysis of the pilot project are provided, attempting to answer the question as to whether QR codes are worth the effort for libraries.

  7. Saphyr: a code system from reactor design to reference calculations

    International Nuclear Information System (INIS)

    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels

  8. Saphyr: a code system from reactor design to reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)

    2003-07-01

    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.

  9. FARO base case post-test analysis by COMETA code

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)

    1995-09-01

    The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.

  10. Severe accident analysis code Sampson for impact project

    Energy Technology Data Exchange (ETDEWEB)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh [Nuclear Power Engineering Corporation, Advanced Simulation Systems Dept., Tokyo (Japan)

    2001-07-01

    Four years of the IMPACT project Phase 1 (1994-1997) had been completed with financial sponsorship from the Japanese government's Ministry of Economy, Trade and Industry. At the end of the phase, demonstration simulations by combinations of up to 11 analysis modules developed for severe accident analysis in the SAMPSON Code were performed and physical models in the code were verified. The SAMPSON prototype was validated by TMI-2 and Phebus-FP test analyses. Many of empirical correlation and conventional models have been replaced by mechanistic models during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. (author)

  11. Methods and computer codes for nuclear systems calculations

    Indian Academy of Sciences (India)

    B P Kochurov; A P Knyazev; A Yu Kwaretzkheli

    2007-02-01

    Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.

  12. Improvement of JRR-4 core management code system

    Energy Technology Data Exchange (ETDEWEB)

    Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N. [Department of Research Reactor, Tokai Research Establishment, Japan Atomic Energy Institute, Tokai, Ibaraki (Japan)

    2000-10-01

    In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

  13. Application bar-code system for solid radioactive waste management

    International Nuclear Information System (INIS)

    Solid radioactive wastes are generated from the post-irradiated fuel examination facility, the irradiated material examination facility, the research reactor, and the laboratories at KAERI. A bar-code system for a solid radioactive waste management of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by bar-code system

  14. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  15. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  16. Computer codes for the analysis of flask impact problems

    International Nuclear Information System (INIS)

    This review identifies typical features of the design of transportation flasks and considers some of the analytical tools required for the analysis of impact events. Because of the complexity of the physical problem, it is unlikely that a single code will adequately deal with all the aspects of the impact incident. Candidate codes are identified on the basis of current understanding of their strengths and limitations. It is concluded that the HONDO-II, DYNA3D AND ABAQUS codes which ar already mounted on UKAEA computers will be suitable tools for use in the analysis of experiments conducted in the proposed AEEW programme and of general flask impact problems. Initial attention should be directed at the DYNA3D and ABAQUS codes with HONDO-II being reserved for situations where the three-dimensional elements of DYNA3D may provide uneconomic simulations in planar or axisymmetric geometries. Attention is drawn to the importance of access to suitable mesh generators to create the nodal coordinate and element topology data required by these structural analysis codes. (author)

  17. Demonstration study on shielding safety analysis code (8)

    International Nuclear Information System (INIS)

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A 3He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, 3He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

  18. Optical System Design For High Speed Bar Code Scanning

    Science.gov (United States)

    Hellekson, Ronald; Reddersen, Brad; Campbell, Scott

    1987-04-01

    Spectra-Physics recently introduced the Model 750 SL scanner for use in the European point-of-sale market, to meet the European requirement for a scanner of less than 13 cm height. The model 750 SL uses a higher density computer designed scan pattern with a retrodirective collection system to scan and detect UPC, EAN, and JAN bar codes. The scanner "reads" these bar codes in such a way that the user need not precisely align the bar code symbol with respect to the window in the scanner even at package speeds up to 100 inches per second. By using a unique geometrical arrangement of mirrors, a polygonal mirror assembly, and a custom-designed plastic bifocal lens, a design was developed to meet these requirements. This paper describes the design of this new low cost scanner, the use of computer-aided design in the development of this scanner, and some observations on the future of bar code scanning.

  19. Programme Code for Projecting of WDM Fiber Optic Sensor Systems

    Directory of Open Access Journals (Sweden)

    R. Probstner

    1993-04-01

    Full Text Available Wavelength division multiplex (WDM offers a potentially powerful technique for use within optical fibre sensor systems. The paper deals with short description of methodology and a programme code for WDM fiber optic sensor system projecting with use of CAD.

  20. 基于程序控制流图源代码相似程度分析系统%Program Control Flow Graph Based on the Similarity of Source Code Analysis System

    Institute of Scientific and Technical Information of China (English)

    陈新

    2013-01-01

      源代码相似程度分析在软件工程和计算机教学等领域都有重要的应用。软件工程领域的源代码盗窃和著作权纠纷仲裁,计算机教学领域的学生作业作弊分析都需要源代码相似程度的分析。良好的源代码相似程序分析软件还可以应用于相似代码聚类和搜索引擎的源代码搜索领域。尽管源代码相似程度分析问题由来已久,但是这个问题并没有令人十分满意和惊喜的研究结果。源代码有其特殊结构,使用传统的纯文本相似度分析显然是不合适的。将首先介绍这个问题的研究历史和进展,简单分析这个问题的难点所在,继而介绍一个新的基于程序控制流图分析的源代码相似程度分析系统,并给出其算法和实现细节。文章最后将分析这个方法的优劣所在,讨论这个方法的进一步改进方向。%It is very important to detect source code similarity in the field of both software engineering and computer science education. Source code stealing and copyright dispute in software engineering, as well as plagiarism detection of student assignment in programming course, call for automation of source code similarity detection. Reliable software to detect source code similarity also helps in the field of source code clustering and source code searching in wide range. Though such problem rises along with the invention of program language, there is not satisfying research result up till now. Source code has its particular structure, making it improper to use traditional pure text similarity detection method over it. This paper first introduce the history and current progress of this problem, analysis the difficulty, then presents a new system for source code similarity detect base on control flow graph analysis, along with the algorithm it uses and the implementation details. In this paper we also discuss the advantage and disadvantage of the method and ways of improve it.

  1. System-Level Genetic Codes Using a Transposable Element-Like Mechanism with Applications to Cancer

    OpenAIRE

    McGowan, John F.

    2000-01-01

    A system-level genetic code is a hypothetical genetic code that exclusively or preferentially codes systems of interacting coadapted parts. System-level genetic codes differ from part-level genetic codes in which each discrete part is coded independently. In general, a system-level genetic code requires coding discrete interacting parts such as organs or proteins in an interdependent way. Changing a single symbol or "gene" in a system-level genetic code affects two or more parts in a coordina...

  2. Development of a frequency-domain coupled neutronic thermal-hydraulic stability analysis code STAC. Verification of thermal-hydraulic part of the code

    International Nuclear Information System (INIS)

    A frequency-domain coupled neutronic thermal-hydraulic stability analysis code STAC is under development in TEPCO Systems Corporation (TEPSYS). The code is composed of the steady-state thermal-hydraulic calculation part and the transfer function calculation part. In the transfer function calculation part, neutronics, fuel heat conduction and thermal-hydraulics models are implemented. In this paper, the thermal hydraulic part of the code is focused on. A basic framework of the code is learned from NUFREQ-NP code. The basic equations are almost the same, but many modifications are conducted. The major modifications in the thermal-hydraulic part are 1) introduction of the finite difference scheme in spatial discretization to simplify the code structure and facilitate the modification of the code and 2) consideration of perturbation in almost all empirical correlations. The STAC code is validated using steady-state pressure drop and void fraction data which were measured in NUPEC full bundle test. A good agreement between predicted and measured values is shown thus the steady-state calculation part of the code is well validated. Then, the STAC code is validated using stability threshold power measurement test data. The stability threshold power is calculated by STAC and the predicted and measured values are compared. Those values agree well. Predicted and measured resonance frequencies are also compared and good agreement is observed. (author)

  3. Validation of system codes for plant application on selected experiments

    Energy Technology Data Exchange (ETDEWEB)

    Koch, Marco K.; Risken, Tobias; Agethen, Kathrin; Bratfisch, Christoph [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2016-05-15

    For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.

  4. Validation of system codes for plant application on selected experiments

    International Nuclear Information System (INIS)

    For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.

  5. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    International Nuclear Information System (INIS)

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), University of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  6. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    Energy Technology Data Exchange (ETDEWEB)

    Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  7. Chaos Many-Body Engine v03: A new version of code C# for chaos analysis of relativistic many-body systems with reactions

    Science.gov (United States)

    Grossu, I. V.; Besliu, C.; Jipa, Al.; Felea, D.; Esanu, T.; Stan, E.; Bordeianu, C. C.

    2013-04-01

    In this paper we present a new version of the Chaos Many-Body Engine C# application (Grossu et al. 2012 [1]). In order to benefit from the latest technological advantages, we migrated the application from .Net Framework 2.0 to .Net Framework 4.0. New tools were implemented also. Trying to estimate the particle interactions dependence on initial conditions, we considered a new distance, which takes into account only the structural differences between two systems. We used this distance for implementing the “Structural Lyapunov” function. We propose also a new precision test based on temporal reversed simulations. New version program summaryProgram title: Chaos Many-Body Engine v03 Catalogue identifier: AEGH_v3_0 Program summary URL: http://cpc.cs.qub.ac.uk/summaries/AEGH_v3_0.html Program obtainable from: CPC Program Library, Queen’s University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 214429 No. of bytes in distributed program, including test data, etc.: 9512380 Distribution format: tar.gz Programming language: Visual C# .Net 2010 Computer: PC Operating system: .Net Framework 4.0 running on MS Windows RAM: 128 MB Classification: 24.60.Lz, 05.45.a Catalogue identifier of previous version: AEGH_v2_0 Journal reference of previous version: Computer Physics Communications 183 (2012) 1055-1059 Does the new version supersede the previous version?: Yes Nature of problem: Chaos analysis of three-dimensional, relativistic many-body systems with reactions. Solution method: Second order Runge-Kutta algorithm. Implementation of temporal reversed simulation precision test, and “Structural Lyapunov” function. In order to benefit from the advantages involved in the latest technologies (e.g. LINQ Queries [2]), Chaos Many-Body Engine was migrated from .Net Framework 2.0 to .Net Framework 4.0. In addition to existing energy conservation

  8. 76 FR 57982 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-09-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Office of Energy Efficiency and Renewable Energy Building Energy Codes Cost Analysis Correction In notice document 2011-23236 beginning on page 56413 in the issue of Tuesday, September 13, 2011 make the...

  9. A systems neurophysiology approach to voluntary event coding.

    Science.gov (United States)

    Petruo, Vanessa A; Stock, Ann-Kathrin; Münchau, Alexander; Beste, Christian

    2016-07-15

    Mechanisms responsible for the integration of perceptual events and appropriate actions (sensorimotor processes) have been subject to intense research. Different theoretical frameworks have been put forward with the "Theory of Event Coding (TEC)" being one of the most influential. In the current study, we focus on the concept of 'event files' within TEC and examine what sub-processes being dissociable by means of cognitive-neurophysiological methods are involved in voluntary event coding. This was combined with EEG source localization. We also introduce reward manipulations to delineate the neurophysiological sub-processes most relevant for performance variations during event coding. The results show that processes involved in voluntary event coding included predominantly stimulus categorization, feature unbinding and response selection, which were reflected by distinct neurophysiological processes (the P1, N2 and P3 ERPs). On a system's neurophysiological level, voluntary event-file coding is thus related to widely distributed parietal-medial frontal networks. Attentional selection processes (N1 ERP) turned out to be less important. Reward modulated stimulus categorization in parietal regions likely reflecting aspects of perceptual decision making but not in other processes. The perceptual categorization stage appears central for voluntary event-file coding. PMID:27153981

  10. Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Kang, H-S; Jang, M-S; Kim, S-R [NESS, Daejeon (Korea, Republic of); Park, J-M; Kim, K-N [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis.

  11. Methodology for coding the energy emergency management information system. [Facility ID's and energy codes

    Energy Technology Data Exchange (ETDEWEB)

    D' Acierno, J.; Hermelee, A.; Fredrickson, C.P.; Van Valkenburg, K.

    1979-11-01

    The coding methodology for creating facility ID's and energy codes from information existing in EIA data systems currently being mapped into the EEMIS data structure is presented. A comprehensive approach is taken to facilitate implementation of EEMIS. A summary of EIA data sources which will be a part of the final system is presented in a table showing the intersection of 19 EIA data systems with the EEMIS data structure. The methodology for establishing ID codes for EIA sources and the corresponding EEMIS facilities in this table is presented. Detailed energy code translations from EIA source systems to the EEMIS energy codes are provided in order to clarify the transfer of energy data from many EIA systems which use different coding schemes. 28 tables.

  12. Rateless Space Time Block Code for Massive MIMO Systems

    Directory of Open Access Journals (Sweden)

    Ali H. Alqahtani

    2014-01-01

    Full Text Available This paper presents a rateless space time block code (RSTBC for massive MIMO systems. The paper illustrates the basis of rateless space time codes deployments in massive MIMO transmissions over wireless erasure channels. In such channels, data may be lost or is not decodable at the receiver due to a variety of factors such as channel fading, interference, or antenna element failure. We show that RSTBC guarantees the reliability of the system in such cases, even when the data loss rate is 25% or more. In such a highly lossy channel, the conventional fixed-rate codes fail to perform well, particularly when channel state information is not available at the transmitter. Simulation results are provided to demonstrate the BER performance and the spectral efficiency of the proposed scheme.

  13. Organization of Risk Analysis Codes for Living Evaluations (ORACLE)

    International Nuclear Information System (INIS)

    ORACLE (Organization of Risk Analysis Codes for Living Evaluations) is an integration concept for using risk-based information in United States Nuclear Regulatory Commission (USNRC) applications. Portions of ORACLE are being developed at the Idaho Nationale Engineering Laboratory for the USNRC. The ORACLE concept consists of related databases, software, user interfaces, processes, and quality control checks allowing a wide variety of regulatory problems and activities to be addressed using current, updated PRA information. The ORACLE concept provides for smooth transitions between one code and the next without pre- or post-processing. (orig.)

  14. FARO and KROTOS code simulation and analysis at JRC Ispra

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Yerkess, A.; Addabbo, C. [European Commission-Joint Research Centre, Inst. for Systems, Informatics and Safety, 21020 Ispra (Italy)

    1998-01-01

    The paper summarizes relevant results from the pre and post test calculations of fuel coolant interaction and quenching tests performed in the FARO and KROTOS test facilities. The main analytical tools adopted at JRC Ispra are the COMETA and the TEXAS codes. COMETA pre and post test calculations of FARO Test L-20 as well as an application of the code to KROTOS test facility are presented. The analysis provides the need to account for H{sub 2} generation models into the pre-mixing calculations. In addition salient results from the application of TEXAS to FARO and KROTOS tests are shown. (author)

  15. Starfinder a code for crowded stellar fields analysis

    CERN Document Server

    Diolaiti, E; Bonaccini, D; Close, L M; Currie, D; Parmeggiani, G

    1999-01-01

    Starfinder is an IDL code for the deep analysis of stellar fields, designed for well-sampled images with high and low Strehl factor. An important feature is represented by the possibility to measure the anisoplanatic effect in wide-field Adaptive Optics observations and exploit this knowledge to improve the analysis of the observed field. A description of the method and applications to real AO data are presented.

  16. Three-Dimensional Turbomachine-Blade-Row Analysis Code

    Science.gov (United States)

    Glassman, A. J.; Wood, J. R.

    1986-01-01

    Computer program (MERNEW3D) developed that prepares bulk of input data set required for Denton three-dimensional inviscid turbomachine-blade-row analysis code. Denton input generated from minimum of geometry and flow-variable information by using cubic spline curve fits for interpolation and extrapolation. Curve-fitting procedures taken from previously developed and widely used NASA computer program (MERIDL), which performs meridional streamsurface analysis.

  17. LSENS, a general chemical kinetics and sensitivity analysis code for homogeneous gas-phase reactions. 2: Code description and usage

    Science.gov (United States)

    Radhakrishnan, Krishnan; Bittker, David A.

    1994-01-01

    LSENS, the Lewis General Chemical Kinetics Analysis Code, has been developed for solving complex, homogeneous, gas-phase chemical kinetics problems and contains sensitivity analysis for a variety of problems, including nonisothermal situations. This report is part 2 of a series of three reference publications that describe LSENS, provide a detailed guide to its usage, and present many example problems. Part 2 describes the code, how to modify it, and its usage, including preparation of the problem data file required to execute LSENS. Code usage is illustrated by several example problems, which further explain preparation of the problem data file and show how to obtain desired accuracy in the computed results. LSENS is a flexible, convenient, accurate, and efficient solver for chemical reaction problems such as static system; steady, one-dimensional, inviscid flow; reaction behind incident shock wave, including boundary layer correction; and perfectly stirred (highly backmixed) reactor. In addition, the chemical equilibrium state can be computed for the following assigned states: temperature and pressure, enthalpy and pressure, temperature and volume, and internal energy and volume. For static problems the code computes the sensitivity coefficients of the dependent variables and their temporal derivatives with respect to the initial values of the dependent variables and/or the three rate coefficient parameters of the chemical reactions. Part 1 (NASA RP-1328) derives the governing equations describes the numerical solution procedures for the types of problems that can be solved by lSENS. Part 3 (NASA RP-1330) explains the kinetics and kinetics-plus-sensitivity-analysis problems supplied with LSENS and presents sample results.

  18. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  19. Design and Implementation of Malicious Code Analysis System Based on Web%基于网络的恶意代码分析系统设计与实现

    Institute of Scientific and Technical Information of China (English)

    任子亭

    2012-01-01

    针对当前形势下恶意代码攻击的新特点以及现有检测技术存在的缺陷,使得恶意代码的分析检测变得越来越困难,文章设计实现的基于网络的恶意代码分析系统,可以自动地对恶意代码进行快速的动态分析,通过数据进行定量和定性相结合的分析,生成详细的分析报告,实验结果表明该系统可以提高恶意代码的分析效率.%For the new .features of malware attacks in the current situation, and the inadequacy of the existing detection technology, the analysis and detection of malware become more difficult. This paper designs and implements a malicious code analysis web-based system, which can provid a dynamic analysis for malware automatically and quickly, the anlayzing to the data through a combination of quantitative and qualitative analysis into a well-organized report. The experiment shows that it can dramatically improve the analysis efficiency.

  20. Automated face analysis by feature point tracking has high concurrent validity with manual FACS coding.

    Science.gov (United States)

    Cohn, J F; Zlochower, A J; Lien, J; Kanade, T

    1999-01-01

    The face is a rich source of information about human behavior. Available methods for coding facial displays, however, are human-observer dependent, labor intensive, and difficult to standardize. To enable rigorous and efficient quantitative measurement of facial displays, we have developed an automated method of facial display analysis. In this report, we compare the results with this automated system with those of manual FACS (Facial Action Coding System, Ekman & Friesen, 1978a) coding. One hundred university students were videotaped while performing a series of facial displays. The image sequences were coded from videotape by certified FACS coders. Fifteen action units and action unit combinations that occurred a minimum of 25 times were selected for automated analysis. Facial features were automatically tracked in digitized image sequences using a hierarchical algorithm for estimating optical flow. The measurements were normalized for variation in position, orientation, and scale. The image sequences were randomly divided into a training set and a cross-validation set, and discriminant function analyses were conducted on the feature point measurements. In the training set, average agreement with manual FACS coding was 92% or higher for action units in the brow, eye, and mouth regions. In the cross-validation set, average agreement was 91%, 88%, and 81% for action units in the brow, eye, and mouth regions, respectively. Automated face analysis by feature point tracking demonstrated high concurrent validity with manual FACS coding.

  1. RHR system reliability analysis of Krsko NPP

    International Nuclear Information System (INIS)

    In this paper Systems reliability analysis is applied to residual heat Removal System in Krsko NPP. Fault tree method is used. Qualitative analysis of the fault tree was made using FTAP-2 computer code, and quantitative using IMPORT code. results are evaluated and their possible application is given. (author)

  2. MMA, A Computer Code for Multi-Model Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Eileen P. Poeter and Mary C. Hill

    2007-08-20

    This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations.

  3. A dual-sided coded-aperture radiation detection system

    Science.gov (United States)

    Penny, R. D.; Hood, W. E.; Polichar, R. M.; Cardone, F. H.; Chavez, L. G.; Grubbs, S. G.; Huntley, B. P.; Kuharski, R. A.; Shyffer, R. T.; Fabris, L.; Ziock, K. P.; Labov, S. E.; Nelson, K.

    2011-10-01

    We report the development of a large-area, mobile, coded-aperture radiation imaging system for localizing compact radioactive sources in three dimensions while rejecting distributed background. The 3D Stand-Off Radiation Detection System (SORDS-3D) has been tested at speeds up to 95 km/h and has detected and located sources in the millicurie range at distances of over 100 m. Radiation data are imaged to a geospatially mapped world grid with a nominal 1.25- to 2.5-m pixel pitch at distances out to 120 m on either side of the platform. Source elevation is also extracted. Imaged radiation alarms are superimposed on a side-facing video log that can be played back for direct localization of sources in buildings in urban environments. The system utilizes a 37-element array of 5×5×50 cm 3 cesium-iodide (sodium) detectors. Scintillation light is collected by a pair of photomultiplier tubes placed at either end of each detector, with the detectors achieving an energy resolution of 6.15% FWHM (662 keV) and a position resolution along their length of 5 cm FWHM. The imaging system generates a dual-sided two-dimensional image allowing users to efficiently survey a large area. Imaged radiation data and raw spectra are forwarded to the RadioNuclide Analysis Kit (RNAK), developed by our collaborators, for isotope ID. An intuitive real-time display aids users in performing searches. Detector calibration is dynamically maintained by monitoring the potassium-40 peak and digitally adjusting individual detector gains. We have recently realized improvements, both in isotope identification and in distinguishing compact sources from background, through the installation of optimal-filter reconstruction kernels.

  4. The Facial Expression Coding System (FACES): Development, Validation, and Utility

    Science.gov (United States)

    Kring, Ann M.; Sloan, Denise M.

    2007-01-01

    This article presents information on the development and validation of the Facial Expression Coding System (FACES; A. M. Kring & D. Sloan, 1991). Grounded in a dimensional model of emotion, FACES provides information on the valence (positive, negative) of facial expressive behavior. In 5 studies, reliability and validity data from 13 diverse…

  5. Development of BERMUDA: a radiation transport code system, 1

    International Nuclear Information System (INIS)

    A radiation transport code system BERMUDA has been developed for one-, two- and three-dimensional geometries. The time-independent transport equation is numerically solved using a direct integration method in a multigroup model, to obtain spatial, angular and energy distributions of neutron, gamma rays or adjoint neutron flux. As to group constants, a library with an any structure of energy groups is capable to be produced from a data base JSSTDL, or by a processing code PROF-GROUCH-G/B, selecting objective nuclear data through a retrieval system EDFSRS. Validity of the present code system has been tested by analyzing the shielding benchmark experiments. The test has shown that accurate results are obtainable with this system especially in deep penetration calculation. Described are the devised calculation method and the results of validity tests. Input data specification, job control languages and output data are also described as a user's manual for the following four neutron transport codes: BERMUDA-1DN : sphere, slab(S20), BERMUDA-2DN : cylinder (S8), BERMUDA-2DN-S16 : cylinder (S16), and BERMUDA-3DN : rectangular parallelpiped (S8). (J.P.N.)

  6. AnDa:a Dynamic Analysis System for Malicious Code%AnDa:恶意代码动态分析系统

    Institute of Scientific and Technical Information of China (English)

    任伟; 柳坤; 周金

    2014-01-01

    近年来,移动终端崛起迅速,其功能已扩展到商务应用等领域,与用户的隐私、财产等信息关系紧密。静态监控已无法满足人们对应用软件安全使用的需求,采用动态监控沙盒分析可以实时监控应用程序,具有速度快、准确性好、安全性高、可行性强的特性。针对Android平台下恶意软件在后台获取用户隐私信息,如获取用户数据并发送到网络端、拦截和窥探用户电话和短信等问题,提出一套采用动态检测沙盒分析技术记录Android恶意软件敏感行为的方案及系统--AnDa,详细描述了该系统总体设计和关键技术,实现了对访问电话、短信、位置信息、手机SIM卡信息等行为的实时监控,并在虚拟机和实体机上测试了AnDa系统。该作品采用动态监控沙盒分析技术,实现了在Android平台下软件动态监控和行为分析,并且实现了对Android框架层API的Java Method Hook和常见的恶意软件特征的有效监控。它可以在Android 4.0以上的设备上使用,可以根据监控到的应用软件恶意行为信息,判定所属恶意软件的类型,使得更加迅速发现新型病毒和更加隐蔽的病毒模型,从而更好地保护手机以及个人重要的数据,极大地提高了安全性。%Recently, mobile terminals have been extended to business applications rapidly, and have been more closely related to user privacy and property. As static monitoring cannot guarantee software security, the analysis of dynamic monitoring sandbox can realize real-time monitoring in a faster, more accurate, safer, and high feasible manner. The problem of privacy leakage exists in Android platform malware, such as accessing user data and exposing them to networks, or intercepting and spying on phone calls and short text messages. Thus, this article proposes a solution system called AnDa, which records sensitive behavior of Android malwares using dynamic

  7. Analysis of LOCA experiments with RELAP4J code

    International Nuclear Information System (INIS)

    The results of analysis with RELAP4J Code are presented for two typical experiments of cold leg break (Runs 413 and 312), in the ROSA-II (Rig of Safety Assessment II) test program. The objectives of analysis are to evaluate validity of the RELAP4J Code, to improve analytical models and to get a better understanding of experimental phenomena. The two tests were performed under actual reactor initial pressure and temperature, in the respective different LPCI locations. Typical factors influencing the pressure history were examined analytically. In conclusion, the predictions of macroscopic-hydraulic phenomena such as pressure transient in each location are good, and the predictions of microscopic-hydraulic phenomena such as steam-water slip velocity, multi-dimentional flow in plenums or core, quenching velocity, cooling of fuel rods by small coolant flow are not good. Experimental phenomena not clarified yet with test data are predicted with the analysis. (author)

  8. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

    International Nuclear Information System (INIS)

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files

  9. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.

  10. Analysis of error performance on Turbo coded FDPIM

    Institute of Scientific and Technical Information of China (English)

    ZHU Yin-bing; WANG Hong-Xing; ZHANG Tie-Ying

    2008-01-01

    Due to variable symbol length of digital pulse interval modulation(DPIM), it is difficult to analyze the error performances of Turbo ceded DPIM. To solve this problem, a fixed-length digital pulse interval modulation(FDPIM) method is provided.The FDPIM modulation structure is introduced. The packet error rates of uncoded FDPIM are analyzed and compared with that of DPIM. Bit error rates of Turbo coded FDPIM are simulated based on three kinds of analytical models under weak turbulence channel. The results show that packet error rate of uncoded FDPIM is inferior to that of uncoded DPIM.However, FDPIM is easy to be implemented and easy to be combined, with Turbo code for soft-decision because of its fixed length. Besides, the introduction of Turbo code in this modulation can decrease the average power about 10 dBm,which means that it can improve the error performance of the system effectively.

  11. Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

    International Nuclear Information System (INIS)

    To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal-hydraulic parameters are predicted by the sub-channel code COBRA-SC. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)

  12. MMA, A Computer Code for Multi-Model Analysis

    Science.gov (United States)

    Poeter, Eileen P.; Hill, Mary C.

    2007-01-01

    This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will

  13. Application of CASMO-3/MASTER Code System to the OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    You, Guk Jong; Sim, Jung Hoon; Kim, Han Gon [KHN, Daejeon (Korea, Republic of)

    2007-10-15

    MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), which was developed by KAERI, is the nuclear design code having the capability of static core design, transient core analysis and operational support. And CASMO-3, which is a fuel assembly burnup program, is the lattice calculation code to generate cross sections for core design code. To validate the core design of APR1400 CASMO- 3/MASTER codes have been selected as independent code system. The core design of APR1400, however, is in progress and the final design data and analysis results are not produced. Therefore, OPR1000, which has sufficient information, is selected as a reference plant to demonstrate the performance of CASMO-3/MASTER code package. This demonstration has been performed using design data of UCN no.4 Cycle1 and the results are compared to Nuclear Design Report(NDR) of the UCN no.4 Cycle 1. The performance of the code package is verified through uncertainty quantification according to the uncertainty evaluation report written by KAERI.

  14. Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

    International Nuclear Information System (INIS)

    A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

  15. Environmental performance of green building code and certification systems.

    Science.gov (United States)

    Suh, Sangwon; Tomar, Shivira; Leighton, Matthew; Kneifel, Joshua

    2014-01-01

    We examined the potential life-cycle environmental impact reduction of three green building code and certification (GBCC) systems: LEED, ASHRAE 189.1, and IgCC. A recently completed whole-building life cycle assessment (LCA) database of NIST was applied to a prototype building model specification by NREL. TRACI 2.0 of EPA was used for life cycle impact assessment (LCIA). The results showed that the baseline building model generates about 18 thousand metric tons CO2-equiv. of greenhouse gases (GHGs) and consumes 6 terajoule (TJ) of primary energy and 328 million liter of water over its life-cycle. Overall, GBCC-compliant building models generated 0% to 25% less environmental impacts than the baseline case (average 14% reduction). The largest reductions were associated with acidification (25%), human health-respiratory (24%), and global warming (GW) (22%), while no reductions were observed for ozone layer depletion (OD) and land use (LU). The performances of the three GBCC-compliant building models measured in life-cycle impact reduction were comparable. A sensitivity analysis showed that the comparative results were reasonably robust, although some results were relatively sensitive to the behavioral parameters, including employee transportation and purchased electricity during the occupancy phase (average sensitivity coefficients 0.26-0.29). PMID:24483287

  16. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-03-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  17. ADAPTIVE ERROR-LIMITING METHOD SUITABLEFOR THE WALSH CODE SHUTTING MULTIPLEXING IN THE MINE MONITOR SYSTEM

    Institute of Scientific and Technical Information of China (English)

    ZhuLiping

    1996-01-01

    Through the analysis for the process of Walsh modulation and demodulation, the adaptive error-limiting method suitable for the Walsh code shutting multiplexing in the mine monitor system is advanced in this article. It is proved by theoretical analysis and circuit experiments that this method is easy to carry out and can not onlyimprove the quality of information transmission but also meet the requirement of thesystem patrol test time without the increasement of system investment.

  18. An Algorithm for Constructing All Families of Codes of Arbitrary Requirement in an OCDMA System

    OpenAIRE

    Lu, Xiang; Chen, Jiajia; He, Sailing

    2006-01-01

    A novel code construction algorithm is presented to find all the possible code families for code reconfiguration in an OCDMA system. The algorithm is developed through searching all the complete subgraphs of a constructed graph. The proposed algorithm is flexible and practical for constructing optical orthogonal codes (OOCs) of arbitrary requirement. Simulation results show that one should choose an appropriate code length in order to obtain sufficient number of code families for code reconfi...

  19. CANDU safety analysis system establishment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Rhee, B. W.; Park, J. H.; Kim, H. T.; Choi, H. B.; Shim, J. I.; Yoon, C.; Yang, M. K

    2002-03-01

    To develop CANDU safety analysis system, methodology, and assessment technology, GAIs from CNSC and GSIs drived by IAEA are summarized. Furthermore, the following safety items are investigated in the present study. - It is intended to secure credibility of the void reactivity in the stage of nuclear design and analysis. The measurement data concerned with the void reactivity were reviewed and used to assess the physics code such as POWDERPUFS-V/RFSP, and the lattice code such as WIMS-AECL and MCNP-4B. - Reviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc. were examined. - The development of 3D CFD transient analysis model has been performed to predict local subcooling of the moderator in the vicinity of Calandria tubes in a CANDU-6 reactor in the case of Large LOCA transient. - The trip coverage analysis methodology based on CATHENA code is developed. The simulation of real plant transient showed good agreement. The trip coverage map was generated successfully for two typical depressurization and pressurization event. - The multi-dimensional analysis methodology for hydrogen distribution and hydrogen burning phenomena in PHWR containment is developed using GOTHIC code. The multi-dimensional analysis predicts the local hydrogen behaviour compared to the lumped parameter model.

  20. Performance Analysis for Dispensing Mechanism of Active Code

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The dispensing mechanism of active code is a key technology in an active network. Conventional capsule and programmable switch approaches have their own shortcomings. The DCCAN(distributed code caching for active network) mechanism presented in this paper overcomes these shortcomings. In this paper, capsule and programmable switch approaches are introduced, and their shortcomings are analyzed. The principle of the DCCAN mechanism is described. The theory analysis in transmit width based on the DCCAN mechanism and capsule approach are described. The theory analysis shows that the DCCAN mechanism has many good characteristics and can improve the efficiency of an active network. Key factors which affect the transmit width based on the DCCAN mechanism are discussed. The using condition of the DCCAN mechanism is also discussed.

  1. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

    International Nuclear Information System (INIS)

    This report presents the results of the COBRA-SFS (Spent Fuel Storage) computer code validation effort. COBRA-SFS, while refined and specialized for spent fuel storage system analyses, is a lumped-volume thermal-hydraulic analysis computer code that predicts temperature and velocity distributions in a wide variety of systems. Through comparisons of code predictions with spent fuel storage system test data, the code's mathematical, physical, and mechanistic models are assessed, and empirical relations defined. The six test cases used to validate the code and code models include single-assembly and multiassembly storage systems under a variety of fill media and system orientations and include unconsolidated and consolidated spent fuel. In its entirety, the test matrix investigates the contributions of convection, conduction, and radiation heat transfer in spent fuel storage systems. To demonstrate the code's performance for a wide variety of storage systems and conditions, comparisons of code predictions with data are made for 14 runs from the experimental data base. The cases selected exercise the important code models and code logic pathways and are representative of the types of simulations required for spent fuel storage system design and licensing safety analyses. For each test, a test description, a summary of the COBRA-SFS computational model, assumptions, and correlations employed are presented. For the cases selected, axial and radial temperature profile comparisons of code predictions with test data are provided, and conclusions drawn concerning the code models and the ability to predict the data and data trends. Comparisons of code predictions with test data demonstrate the ability of COBRA-SFS to successfully predict temperature distributions in unconsolidated or consolidated single and multiassembly spent fuel storage systems

  2. A guide to the AUS modular neutronics code system

    International Nuclear Information System (INIS)

    A general description is given of the AUS modular neutronics code system, which may be used for calculations of a very wide range of fission reactors, fusion blankets and other neutron applications. The present system has cross-section libraries derived from ENDF/B-IV and includes modules which provide for lattice calculations, one-dimensional transport calculations, and one, two, and three-dimensional diffusion calculations, burnup calculations and the flexible editing of results. Details of all system aspects of AUS are provided but the major individual modules are only outlined. Sufficient information is given to enable other modules to be added to the system

  3. Development of the Verification and Validation Matrix for Safety Analysis Code SPACE

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yo Han; Ha, Sang Jun; Yang, Chang Keun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Korea Electric Power Research Institute (KEPRI) has been developed the safety analysis code, called as SPACE (Safety and Performance Analysis CodE for Nuclear Power Plant), for typical pressurized water reactors (PWR). Current safety analysis codes were conducted from foreign vendors, such as Westinghouse Electric Corp., ABB Combustion Engineering Inc., Kraftwerk Union, etc. Considering the conservatism and inflexibility of the foreign code systems, it is difficult to expand the application areas and analysis scopes. To overcome the mentioned problems KEPRI has launched the project to develop the native safety analysis code with Korea Power Engineering Co.(KOPEC), Korea Atomic Energy Research Inst.(KAERI), Korea Nuclear Fuel(KNF), and Korea Hydro and Nuclear Power Co.(KHNP) under the funding of Ministry of Knowledge Economy (MKE). As a result of the project, the demo-version of SPACE has been released in July 2009. As an advance preparation of the next step, KEPRI and colleagues have developed the verification and validation (V and V) matrix for SPACE. To develop the matrix, the preceding studies and experiments were reviewed. After mature consideration, the V and V matrix has been developed and the experiment plans were designed for the next step to compensate the lack of data.

  4. Comparative study of Barcode, QR-code and RFID System

    OpenAIRE

    Trupti Lotlikar; Rohan Kankapurkar; Anand Parekar; Akshay Mohite

    2013-01-01

    Wireless sensors are standard measurement tools equipped with transmitters to convert signals from process control instruments into a radio transmission. The radio signal is interpreted by a receiver which then converts the wireless signal to a specific, desired output, such as an analog current or data analysis via computer software. The paper gives a brief on wireless sensors and their types like Barcode, QR code, RFID along with their characteristics and working components. The Barcode is ...

  5. HSI colour-coded analysis of scattered light of single plasmonic nanoparticles

    Science.gov (United States)

    Zhou, Jun; Lei, Gang; Zheng, Lin Ling; Gao, Peng Fei; Huang, Cheng Zhi

    2016-06-01

    Single plasmonic nanoparticles (PNPs) analysis with dark-field microscopic imaging (iDFM) has attracted much attention in recent years. The ability for quantitative analysis of iDFM is critical, but cumbersome, for characterizing and analyzing the scattered light of single PNPs. Here, a simple automatic HSI colour coding method is established for coding dark-field microscopic (DFM) images of single PNPs with localized surface plasmon resonance (LSPR) scattered light, showing that hue value in the HSI system can realize accurate quantitative analysis of iDFM and providing a novel approach for quantitative chemical and biochemical imaging at the single nanoparticle level.Single plasmonic nanoparticles (PNPs) analysis with dark-field microscopic imaging (iDFM) has attracted much attention in recent years. The ability for quantitative analysis of iDFM is critical, but cumbersome, for characterizing and analyzing the scattered light of single PNPs. Here, a simple automatic HSI colour coding method is established for coding dark-field microscopic (DFM) images of single PNPs with localized surface plasmon resonance (LSPR) scattered light, showing that hue value in the HSI system can realize accurate quantitative analysis of iDFM and providing a novel approach for quantitative chemical and biochemical imaging at the single nanoparticle level. Electronic supplementary information (ESI) available: Experimental section and additional figures. See DOI: 10.1039/c6nr01089j

  6. Impact of Different Spreading Codes Using FEC on DWT Based MC-CDMA System

    OpenAIRE

    Masum, Saleh; Kabir, M. Hasnat; Islam, Md. Matiqul; Shams, Rifat Ara; Ullah, Shaikh Enayet

    2012-01-01

    The effect of different spreading codes in DWT based MC-CDMA wireless communication system is investigated. In this paper, we present the Bit Error Rate (BER) performance of different spreading codes (Walsh-Hadamard code, Orthogonal gold code and Golay complementary sequences) using Forward Error Correction (FEC) of the proposed system. The data is analyzed and is compared among different spreading codes in both coded and uncoded cases. It is found via computer simulation that the performance...

  7. Load Flow Analysis Using Real Coded Genetic Algorithm

    Directory of Open Access Journals (Sweden)

    Himakar Udatha

    2014-02-01

    Full Text Available This paper presents a Real Coded Genetic Algorithm (RCGA for finding the load flow solution of electrical power systems. The proposed method is based on the minimization of the real and reactive power mismatches at various buses. The traditional methods such as Gauss-Seidel method and Newton-Raphson (NR method have certain drawbacks under abnormal operating condition. In order to overcome these problems, the load flow solution based on Real Coded Genetic Algorithm (RCGA is presented in this paper. Two cross over techniques, Arithmetic crossover and heuristic crossover are used to solve the power flow problem. The proposed method is applied for 3-bus, 5-bus and 6-bus systems and the results are presented.

  8. Analysis of airborne radiometric data. Volume 2. Description, listing, and operating instructions for the code DELPHI/MAZAS. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Sperling, M.; Shreve, D.C.

    1978-12-01

    The computer code DELPHI is an interactive English language command system for the analysis of airborne radiometric data. The code includes modules for data reduction, data simulation, time filtering, data adjustment and graphical presentation of the results. DELPHI is implemented in FORTRAN on a DEC-10 computer. This volume gives a brief set of operations instructions, samples of the output obtained from hard copies of the display on a Tektronix terminal and finally a listing of the code.

  9. Analysis of airborne radiometric data. Volume 2. Description, listing, and operating instructions for the code DELPHI/MAZAS. Final report

    International Nuclear Information System (INIS)

    The computer code DELPHI is an interactive English language command system for the analysis of airborne radiometric data. The code includes modules for data reduction, data simulation, time filtering, data adjustment and graphical presentation of the results. DELPHI is implemented in FORTRAN on a DEC-10 computer. This volume gives a brief set of operations instructions, samples of the output obtained from hard copies of the display on a Tektronix terminal and finally a listing of the code

  10. Photovoltaic power systems and the National Electrical Code: Suggested practices

    Energy Technology Data Exchange (ETDEWEB)

    Wiles, J. [New Mexico State Univ., Las Cruces, NM (United States). Southwest Technology Development Inst.

    1996-12-01

    This guide provides information on how the National Electrical Code (NEC) applies to photovoltaic systems. The guide is not intended to supplant or replace the NEC; it paraphrases the NEC where it pertains to photovoltaic systems and should be used with the full text of the NEC. Users of this guide should be thoroughly familiar with the NEC and know the engineering principles and hazards associated with electrical and photovoltaic power systems. The information in this guide is the best available at the time of publication and is believed to be technically accurate; it will be updated frequently. Application of this information and results obtained are the responsibility of the user.

  11. Channel estimation for physical layer network coding systems

    CERN Document Server

    Gao, Feifei; Wang, Gongpu

    2014-01-01

    This SpringerBrief presents channel estimation strategies for the physical later network coding (PLNC) systems. Along with a review of PLNC architectures, this brief examines new challenges brought by the special structure of bi-directional two-hop transmissions that are different from the traditional point-to-point systems and unidirectional relay systems. The authors discuss the channel estimation strategies over typical fading scenarios, including frequency flat fading, frequency selective fading and time selective fading, as well as future research directions. Chapters explore the performa

  12. Demonstration study on shielding safety analysis code. 7

    International Nuclear Information System (INIS)

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) To improve the detection sensitivity of pulse neutron measurement, two neutron detectors and some electronic circuits are added to the system constructed last year. (2) To estimate the neutron dose at the distant point from the facility instead of the commercialized rem-counter, a 3He detector with paraffin moderator is equipped to the system. (3) Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility was measured in the distance up to 300 m. The results show that the time structure of pulsed neutrons almost disappears at the further points than 150 m. (4) In the distance from 90 m to 300 m ordinal total counting method without gate pulse are applied to detect the neutrons. (5) The experimental results of space dependency up to 300 m is fitted fairly well by the Gui's response function. (author)

  13. Neural map formation and sensory coding in the vomeronasal system.

    Science.gov (United States)

    Brignall, Alexandra C; Cloutier, Jean-François

    2015-12-01

    Sensory systems enable us to encode a clear representation of our environment in the nervous system by spatially organizing sensory stimuli being received. The organization of neural circuitry to form a map of sensory activation is critical for the interpretation of these sensory stimuli. In rodents, social communication relies strongly on the detection of chemosignals by the vomeronasal system, which regulates a wide array of behaviours, including mate recognition, reproduction, and aggression. The binding of these chemosignals to receptors on vomeronasal sensory neurons leads to activation of second-order neurons within glomeruli of the accessory olfactory bulb. Here, vomeronasal receptor activation by a stimulus is organized into maps of glomerular activation that represent phenotypic qualities of the stimuli detected. Genetic, electrophysiological and imaging studies have shed light on the principles underlying cell connectivity and sensory map formation in the vomeronasal system, and have revealed important differences in sensory coding between the vomeronasal and main olfactory system. In this review, we summarize the key factors and mechanisms that dictate circuit formation and sensory coding logic in the vomeronasal system, emphasizing differences with the main olfactory system. Furthermore, we discuss how detection of chemosignals by the vomeronasal system regulates social behaviour in mice, specifically aggression. PMID:26329476

  14. Pairwise codeword error probability for coded atmospheric optical communication systems

    Institute of Scientific and Technical Information of China (English)

    HAN Jia-jia; RONG Jian; ZHONG Xiao-chun

    2006-01-01

    To study the performance of various error-control coding schemes,exact expressions and upper bounds on the pairwise codeword error probability(PEP)for several modulation schemes(OOK,SC-BPSK,BPPM)used in atmospheric optical communication systems are derived.To simplify the computation,this research was under the assumption of weak turbulence.Moreover,by simulation of expressions,the performances of PEP in different modulation schemes are compared and the best one of them is given.

  15. V.S.O.P.('94) computer code system for reactor physics and fuel cycle simulation

    International Nuclear Information System (INIS)

    V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories and temporary in-depth research. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to HTR's). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The storage requirement is confined to 17 M-Bytes. The code system has extensively been used for comparison studies of reactors, their fuel cycles, simulation of safety features, developmental research, and reactor assessments. Beside its use in research and development work for the gas cooled High Temperature Reactor the code has succesfully been applied to Light Water Reactors, Heavy Water Reactors, and hybride systems with different moderators. (orig.)

  16. Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning

    International Nuclear Information System (INIS)

    Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs

  17. Demonstration of Emulator-Based Bayesian Calibration of Safety Analysis Codes: Theory and Formulation

    Directory of Open Access Journals (Sweden)

    Joseph P. Yurko

    2015-01-01

    Full Text Available System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here with Markov Chain Monte Carlo (MCMC sampling feasible. This work uses Gaussian Process (GP based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.

  18. Demonstration of emulator-based Bayesian calibration of safety analysis codes: Theory and formulation

    International Nuclear Information System (INIS)

    System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator) construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here with Markov Chain Monte Carlo (MCMC) sampling feasible. This study uses Gaussian Process (GP) based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This 'function factorization' Gaussian Process (FFGP) model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process

  19. Research of Wavelet Based Multicarrier Modulation System with Near Shannon Limited Codes

    Institute of Scientific and Technical Information of China (English)

    ZHANGHaixia; YUANDongfeng; ZHAOFeng

    2005-01-01

    In this paper, by using turbo codes and Low density parity codes (LDPC) as channel correcting code scheme, Wavelet based multicarrier modulation (WMCM) systems are proposed and investigated on different transmission scenarios. The Bit error rate (BER) performance of these two near Shannon limited codes is simulated and compared with various code parameters. Simulated results show that Turbo coded WMCM (TCWMCM) performs better than LDPC coded WMCM (LDPC-CWMCM) on both AWGN and Rayleigh fading channels when these two kinds of codes are of the same code parameters.

  20. Stellarator-specific developments for the systems code PROCESS

    Energy Technology Data Exchange (ETDEWEB)

    Warmer, Felix; Beidler, Craig; Dinklage, Andreas; Feng, Yuehe; Geiger, Joachim; Schauer, Felix; Turkin, Yuriy; Wolf, Robert; Xanthopoulos, Pavlos [Max-Planck-Institut fuer Plasmaphysik, Wendelsteinstrasse 1, D-17491 Greifswald (Germany); Knight, Peter; Ward, David [Culham Centre for Fusion Energy, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)

    2014-07-01

    The ultimate goal of fusion research is to demonstrate the feasibility of economic production of electricity. The most promising concepts to achieve this by magnetic confinement are the Tokamak and the Stellarator. System codes are used to study the general properties of a fusion power plant. Built in a modular way systems codes describe the physical and technical properties of the power plant components. For the Helical Advanced Stellarator (HELIAS) concept modules have been developed in the frame of the existing Tokamak systems code PROCESS. These include: A geometry model based on Fourier coefficients which represent the complex 3-D plasma shape, a divertor model which assumes diffusive cross-field transport and high radiation at the X-point, a coil model which uses a scaling based on the HELIAS design and a transport model which either employs empirical confinement time scalings or sophisticated 1-D collisional and turbulent transport calculations. This approach aims at a direct comparison between Tokamak and Stellarator power plant designs.

  1. Dictionary Learning for Sparse Coding: Algorithms and Convergence Analysis.

    Science.gov (United States)

    Bao, Chenglong; Ji, Hui; Quan, Yuhui; Shen, Zuowei

    2016-07-01

    In recent years, sparse coding has been widely used in many applications ranging from image processing to pattern recognition. Most existing sparse coding based applications require solving a class of challenging non-smooth and non-convex optimization problems. Despite the fact that many numerical methods have been developed for solving these problems, it remains an open problem to find a numerical method which is not only empirically fast, but also has mathematically guaranteed strong convergence. In this paper, we propose an alternating iteration scheme for solving such problems. A rigorous convergence analysis shows that the proposed method satisfies the global convergence property: the whole sequence of iterates is convergent and converges to a critical point. Besides the theoretical soundness, the practical benefit of the proposed method is validated in applications including image restoration and recognition. Experiments show that the proposed method achieves similar results with less computation when compared to widely used methods such as K-SVD. PMID:26452248

  2. Forming Teams for Teaching Programming based on Static Code Analysis

    Directory of Open Access Journals (Sweden)

    Davis Arosemena-Trejos

    2012-03-01

    Full Text Available The use of team for teaching programming can be effective in the classroom because it helps students to generate and acquire new knowledge in less time, but these groups to be formed without taking into account some respects, may cause an adverse effect on the teaching-learning process. This paper proposes a tool for the formation of team based on the semantics of source code (SOFORG. This semantics is based on metrics extracted from the preferences, styles and good programming practices. All this is achieved through a static analysis of code that each student develops. In this way, you will have a record of students with the information extracted; it evaluates the best formation of teams in a given course. The team€™s formations are based on programming styles, skills, pair programming or with leader.

  3. Forming Teams for Teaching Programming based on Static Code Analysis

    CERN Document Server

    Arosemena-Trejos, Davis; Clunie, Clifton

    2012-01-01

    The use of team for teaching programming can be effective in the classroom because it helps students to generate and acquire new knowledge in less time, but these groups to be formed without taking into account some respects, may cause an adverse effect on the teaching-learning process. This paper proposes a tool for the formation of team based on the semantics of source code (SOFORG). This semantics is based on metrics extracted from the preferences, styles and good programming practices. All this is achieved through a static analysis of code that each student develops. In this way, you will have a record of students with the information extracted; it evaluates the best formation of teams in a given course. The team's formations are based on programming styles, skills, pair programming or with leader.

  4. TRAWA, a transient analysis code for water reactions

    International Nuclear Information System (INIS)

    TRAWA is a transient analysis code for water reactors. It solves the two-group neutron diffusion equations simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. At most one-dimensional submodels are used. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermohydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channel and risers with two-phase flow and of pump lines with incompressible flow. Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. Various transients can be calculated by applying external disturbances. The code is extensively supplied by input and output capabilities. TRAWA is written in FORTRAN V for UNIVAC 1108 computer. (author)

  5. SECURITY ANALYSIS OF MOBILE AUTHENTICATION USING QR-CODES

    Directory of Open Access Journals (Sweden)

    Siwon Sung

    2015-12-01

    Full Text Available The QR-Code authentication system using mobile application is easily implemented in a mobile device with high recognition rate without short distance wireless communication support such as NFC. This system has been widely used for physical authentication system does not require a strong level of security. The system also can be implemented at a low cost. However, the system has a vulnerability of tampering or counterfeiting, because of the nature of the mobile application that should be installed on the user’s smart device. In this paper we analyze the vulnerabilities about each type of architectures of the system and discuss the concerns about the implementation aspect to reduce these vulnerabilities.

  6. Development of environmental dose assessment system (EDAS) code of PC version

    Energy Technology Data Exchange (ETDEWEB)

    Taki, Mitsumasa; Kikuchi, Masamitsu; Kobayashi, Hideo; Yamaguchi, Takenori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-05-01

    A computer code (EDAS) was developed to assess the public dose for the safety assessment to get the license of nuclear reactor operation. This code system is used for the safety analysis of public around the nuclear reactor in normal operation and severe accident. This code was revised and composed for personal computer user according to the Nuclear Safety Guidelines reflected the ICRP1990 recommendation. These guidelines are revised by Nuclear Safety Commission on March, 2001, which are 'Weather analysis guideline for the safety assessment of nuclear power reactor', 'Public dose around the facility assessment guideline corresponding to the objective value for nuclear power light water reactor' and 'Public dose assessment guideline for safety review of nuclear power light water reactor'. This code has been already opened for public user by JAERI, and English version code and user manual are also prepared. This English version code is helpful for international cooperation concerning the nuclear safety assessment with JAERI. (author)

  7. User's guide for the GSMP/OCMHD system code

    Energy Technology Data Exchange (ETDEWEB)

    Dennis, C. B.; Berry, G. F.

    1980-12-01

    The Systems Analysis group of the ANL Engineering Division conducts overall system studies for various power plant concepts, utilizing a computer simulation code. Analytical investigations explore a range of possible performance variables, in order to determine the sensitivity of a specific plant design to variation in key system parameters and, ultimately, to establish probable system performance limits. To accomplish this task, a Generalized System Modeling Program (GSMP) has been developed that will analyze and simulate the particular system of interest for any number of different configurations, automatically holding constraints while conducting either sensitivity studies or optimizations. One system investigated, while developing the ANL/GSMP code, is an open-cycle magneto-hydrodynamic (OCMHD) power plant. By linking mathematical models representing these OCMHD power plant components to the executive level GSMP driver the resulting system code, GSMP/OCMHD, can be used to simulate any OCMHD power plant configuration. This report, a user's guide for GSMP/OCMHD, describes the process for setting up an OCMHD configuration, preparing the input defining that configuration, running the computer code and interpreting the results generated.

  8. The ICPC coding system in pharmacy : developing a subset, ICPC-Ph

    NARCIS (Netherlands)

    van Mil, JWF; Brenninkmeijer, R; Tromp, TFJ

    1998-01-01

    The ICPC system is a coding system developed for general medical practice, to be able to code the GP-patient encounters and other actions. Some of the codes can be easily used by community pharmacists to code complaints and diseases in pharmaceutical care practice. We developed a subset of the ICPC

  9. Numerical study of subcooled boiling phenomena using a component analysis code, CUPID

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ba-Ro; Lee, Yeon-Gun [Jeju National University, Jeju (Korea, Republic of)

    2015-10-15

    In this study, a couple of subcooled boiling experiments at high- (> 10 bar) and low-pressure (near atmospheric pressure) conditions are analyzed using a three-dimensional thermal-hydraulic component code, CUPID. And then the analysis results compared with the results using MARS-KS code. Subcooled boiling experiments at high- and low pressure conditions are analyzed using a three dimensional thermal-hydraulic component code, CUPID. The predictions of the CUPID code shows good agreement with Christenses's data and Bartolomey's data obtained at high pressure conditions. Subcooled boiling is encountered in many industrial applications in the power and process industry. In nuclear reactors, under certain conditions, subcooled boiling may be encountered in the core. The movement of bubbles generated by subcooled boiling affect the heat transfer characteristics and the pressure drop of the system. Thus some experimental and analysis using safety codes works have been already performed by previous investigators. It has been reported that the existing safety analysis codes have some weaknesses in predicting subcooled boiling phenomena at low pressure conditions. Thus, it is required to improve the predictive capability of thermal-hydraulic analysis codes on subcooled boiling phenomenon at low-pressure conditions. At low pressure condition, the CUPID code generally is overestimated prediction of the void fraction. Thus, we did selected submodels in the heat partitioning model by sensitivity analysis. Selected submodels of M{sub c}ase 4 are Kocamustafaogullari and Ishii correlation model of active nucleate site density, N' and Fritz correlation model of bubble departure diameter, d{sub Bd} . And then, case 5 - 8 are reanalysis using submodels of M{sub c}ase 4. The calculated void fraction is compared the default CUPID code model to the modified CUPID code model. As a result, average void fraction error was reduced from 0.081 to 0.011 and 0.128 to 0.024, 0

  10. Meanline Analysis of Turbines with Choked Flow in the Object-Oriented Turbomachinery Analysis Code

    Science.gov (United States)

    Hendricks, Eric S.

    2016-01-01

    The Object-Oriented Turbomachinery Analysis Code (OTAC) is a new meanline/streamline turbomachinery modeling tool being developed at NASA GRC. During the development process, a limitation of the code was discovered in relation to the analysis of choked flow in axial turbines. This paper describes the relevant physics for choked flow as well as the changes made to OTAC to enable analysis in this flow regime.

  11. System analysis and design

    International Nuclear Information System (INIS)

    This book deals with information technology and business process, information system architecture, methods of system development, plan on system development like problem analysis and feasibility analysis, cases for system development, comprehension of analysis of users demands, analysis of users demands using traditional analysis, users demands analysis using integrated information system architecture, system design using integrated information system architecture, system implementation, and system maintenance.

  12. Systemizers Are Better Code-Breakers: Self-Reported Systemizing Predicts Code-Breaking Performance in Expert Hackers and Naïve Participants.

    Science.gov (United States)

    Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena

    2016-01-01

    Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders-attention to detail and systemizing-may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings. PMID:27242491

  13. Systemizers Are Better Code-Breakers: Self-Reported Systemizing Predicts Code-Breaking Performance in Expert Hackers and Naïve Participants.

    Science.gov (United States)

    Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena

    2016-01-01

    Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders-attention to detail and systemizing-may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings.

  14. Advanced coding techniques for few mode transmission systems.

    Science.gov (United States)

    Okonkwo, Chigo; van Uden, Roy; Chen, Haoshuo; de Waardt, Huug; Koonen, Ton

    2015-01-26

    We experimentally verify the advantage of employing advanced coding schemes such as space-time coding and 4 dimensional modulation formats to enhance the transmission performance of a 3-mode transmission system. The performance gain of space-time block codes for extending the optical signal-to-noise ratio tolerance in multiple-input multiple-output optical coherent spatial division multiplexing transmission systems with respect to single-mode transmission performance are evaluated. By exploiting the spatial diversity that few-mode-fibers offer, with respect to single mode fiber back-to-back performance, significant OSNR gains of 3.2, 4.1, 4.9, and 6.8 dB at the hard-decision forward error correcting limit are demonstrated for DP-QPSK 8, 16 and 32 QAM, respectively. Furthermore, by employing 4D constellations, 6 × 28Gbaud 128 set partitioned quadrature amplitude modulation is shown to outperform conventional 8 QAM transmission performance, whilst carrying an additional 0.5 bit/symbol.

  15. Performance Analysis of Multi-user Multi Input Multi Output- Interleave-Division Multiple-Access System Employing Turbo Coding with Multi-User Detection over Frequency-Selective Wireless Communication Channel

    Directory of Open Access Journals (Sweden)

    Kuttathatti S. Vishvaksenan

    2011-01-01

    Full Text Available Problem statement: This study presents the performance analysis of multi-user Multi Input Multi Output (MIMO assisted interleave based multiple-access system. In IDMA, different interleavers are used to distinguish users as against different signature sequence in a conventional code-Division Multiple-Access (CDMA scheme. Approach: The basic principle of IDMA is that the interleaver is unique for the users. Results: In this study, we consider that Interleavers are generated independently and randomly. Also the IDMA technique is extended to multi user MIMO IDMA with multi-user detection. At the receiver, OSIC detector is realized using ZF for frequency fading channel to combat MAI and MUI problem. The performance of the system is analyzed for different channel conditions using extensive simulation runs based on Monte Carlo simulation trials. Conclusion: It is shown that the IDMA scheme can achieve near single user performance in situations with very large numbers of users while maintaining very low receiver complexity. It is discerned from the computer simulation results that IDMA outperforms CDMA in frequency selective channel for high load conditions.

  16. Research and implementation of flexible coding system oriented multi-view

    Institute of Scientific and Technical Information of China (English)

    ZHANG Xuhui; ZHANG Xu; NING Ruxin

    2007-01-01

    On the basis of the requirements of a product data management system (PDM) for the flexible coding system,the principle of the flexible coding system oriented multiview is analyzed. Generation and utilization of coding should be associated with the context of the object. The architecture of the flexible coding system oriented multi-view is studied and the implementation class diagram of the system is designed. The system can support the establishment of five types of code segments, provide the tools of flexible defining coding rules and drive the automatic generation of object coding in different views (contexts). On the foundation of the characteristics of the system, coding for parts is taken as a sample to validate and elaborate the flexible coding process of the system.

  17. Analysis of Edge Detection in Bar Code Symbols: An Overview and Open Problems

    Directory of Open Access Journals (Sweden)

    Saša Krešić-Jurić

    2012-01-01

    Full Text Available Accurate edge localization is essential in bar code decoding. Since speckle noise is the most dominant form of noise in laser bar code scanners, it is important to fully understand its effects on edge detection. Starting with the basic statistical properties of speckle patterns, we present stochastic analysis of speckle noise. We derive the autocorrelation function and power spectral density (PSD of the noise in terms of intensity distribution of the scanning beam. We then study the signal-to-noise ratio for signals that result from scanning different configurations of edges. Next, we consider statistical properties of edge localization error caused by speckle noise. We show that the standard deviation of the error is determined by the PSD of the noise and relative positions of edges in a bar code symbol. Based on the analysis presented here, we propose new criteria for system design.

  18. Nexus: A modular workflow management system for quantum simulation codes

    Science.gov (United States)

    Krogel, Jaron T.

    2016-01-01

    The management of simulation workflows represents a significant task for the individual computational researcher. Automation of the required tasks involved in simulation work can decrease the overall time to solution and reduce sources of human error. A new simulation workflow management system, Nexus, is presented to address these issues. Nexus is capable of automated job management on workstations and resources at several major supercomputing centers. Its modular design allows many quantum simulation codes to be supported within the same framework. Current support includes quantum Monte Carlo calculations with QMCPACK, density functional theory calculations with Quantum Espresso or VASP, and quantum chemical calculations with GAMESS. Users can compose workflows through a transparent, text-based interface, resembling the input file of a typical simulation code. A usage example is provided to illustrate the process.

  19. Fuel Behavior Simulation Code FEMAXI-FBR Development for SFR Core Disruptive Accident Analysis

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been developing ASTERIA-FBR code system for SFR core disruptive accident analysis to contribute as a part of the regulation activity for Japanese prototype FBR, MONJU. The ASTERIA-FBR code system consists of detailed fuel behavior analysis module (FEMAXI-FBR), neutronic Monte-Carlo calculation module (GMVP), and thermal hydraulic module (CONCORD). The calculation scope of the ASTERIA-FBR covers the initiating, transitional and post disassembly expansion processes. The FEMAXI-FBR is based on LWR fuel behavior simulation code FEMAXI-6 and modified the material properties and the calculation models under steady state and transient operational condition. The FEMAXI-FBR has been verified in steady state calculations compared with those of SAS-4A code. Furthermore, the code has been validated by French CABRI slow-TOP (E12) and fast-TOP (BI2) transient calculations. Through these verification and validation, good agreement has been obtained with the FP-gas release ratio, the fuel restructuring, the gap width between pellet and cladding, and the fuel pin failure position. (author)

  20. Advanced Error-Control Coding Methods Enhance Reliability of Transmission and Storage Data Systems

    Directory of Open Access Journals (Sweden)

    K. Vlcek

    2003-04-01

    Full Text Available Iterative coding systems are currently being proposed and acceptedfor many future systems as next generation wireless transmission andstorage systems. The text gives an overview of the state of the art initerative decoded FEC (Forward Error-Correction error-control systems.Such systems can typically achieve capacity to within a fraction of adB at unprecedented low complexities. Using a single code requires verylong code words, and consequently very complex coding system. One wayaround the problem of achieving very low error probabilities is turbocoding (TC application. A general model of concatenated coding systemis shown - an algorithm of turbo codes is given in this paper.

  1. A study on the nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    According to current software development and quality assurance trends. It is necessary to develop computer code management system for nuclear programs. For this reason, the project started in 1987. Main objectives of the project are to establish a nuclear computer code management system, to secure software reliability, and to develop nuclear computer code packages. Contents of performing the project in this year were to operate and maintain computer code information system of KAERI computer codes, to develop application tool, AUTO-i, for solving the 1st and 2nd moments of inertia on polygon or circle, and to research nuclear computer code conversion between different machines. For better supporting the nuclear code availability and reliability, assistance from users who are using codes is required. Lastly, for easy reference about the codes information, we presented list of code names and information on the codes which were introduced or developed during this year. (Author)

  2. LSENS, a general chemical kinetics and sensitivity analysis code for gas-phase reactions: User's guide

    Science.gov (United States)

    Radhakrishnan, Krishnan; Bittker, David A.

    1993-01-01

    A general chemical kinetics and sensitivity analysis code for complex, homogeneous, gas-phase reactions is described. The main features of the code, LSENS, are its flexibility, efficiency and convenience in treating many different chemical reaction models. The models include static system, steady, one-dimensional, inviscid flow, shock initiated reaction, and a perfectly stirred reactor. In addition, equilibrium computations can be performed for several assigned states. An implicit numerical integration method, which works efficiently for the extremes of very fast and very slow reaction, is used for solving the 'stiff' differential equation systems that arise in chemical kinetics. For static reactions, sensitivity coefficients of all dependent variables and their temporal derivatives with respect to the initial values of dependent variables and/or the rate coefficient parameters can be computed. This paper presents descriptions of the code and its usage, and includes several illustrative example problems.

  3. Analysis of Doppler Effect on the Pulse Compression of Different Codes Emitted by an Ultrasonic LPS

    Directory of Open Access Journals (Sweden)

    Jorge Morera

    2011-11-01

    Full Text Available This work analyses the effect of the receiver movement on the detection by pulse compression of different families of codes characterizing the emissions of an Ultrasonic Local Positioning System. Three families of codes have been compared: Kasami, Complementary Sets of Sequences and Loosely Synchronous, considering in all cases three different lengths close to 64, 256 and 1,024 bits. This comparison is first carried out by using a system model in order to obtain a set of results that are then experimentally validated with the help of an electric slider that provides radial speeds up to 2 m/s. The performance of the codes under analysis has been characterized by means of the auto-correlation and cross-correlation bounds. The results derived from this study should be of interest to anyone performing matched filtering of ultrasonic signals with a moving emitter/receiver.

  4. Analysis of Doppler effect on the pulse compression of different codes emitted by an ultrasonic LPS.

    Science.gov (United States)

    Paredes, José A; Aguilera, Teodoro; Alvarez, Fernando J; Lozano, Jesús; Morera, Jorge

    2011-01-01

    This work analyses the effect of the receiver movement on the detection by pulse compression of different families of codes characterizing the emissions of an ultrasonic local positioning system. Three families of codes have been compared: Kasami, Complementary Sets of Sequences and Loosely Synchronous, considering in all cases three different lengths close to 64, 256 and 1,024 bits. This comparison is first carried out by using a system model in order to obtain a set of results that are then experimentally validated with the help of an electric slider that provides radial speeds up to 2 m/s. The performance of the codes under analysis has been characterized by means of the auto-correlation and cross-correlation bounds. The results derived from this study should be of interest to anyone performing matched filtering of ultrasonic signals with a moving emitter/receiver.

  5. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (porting). Progress report fiscal 1996

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki [Fujitsu Ltd., Tokyo (Japan); Kawasaki, Nobuo; Tanabe, Hidenobu [and others

    1998-01-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the porting. In this porting part, the porting of reactor safety analysis code RELAP5/MOD3.2 and RELAP5/MOD3.2.1.2, nuclear data processing system NJOY and 2-D multigroup discrete ordinate transport code TWOTRAN-II are described. And also, a survey for the porting of command-driven interactive data analysis plotting program IPLOT are described. In the parallelization part, the parallelization of 2-Dimensional relativistic electromagnetic particle code EM2D, Cylindrical Direct Numerical Simulation code CYLDNS and molecular dynamics code for simulating radiation damages in diamond crystals DGR are described. And then, in the vectorization part, the vectorization of two and three dimensional discrete ordinates simulation code DORT-TORT, gas dynamics analysis code FLOWGR and relativistic Boltzmann-Uehling-Uhlenbeck simulation code RBUU are described. (author)

  6. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (parallelization). Progress report fiscal 1996

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Hideo; Kawai, Wataru; Nemoto, Toshiyuki [Fujitsu Ltd., Tokyo (Japan)] [and others

    1997-12-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the parallelization. In this parallelization part, the parallelization of 2-Dimensional relativistic electromagnetic particle code EM2D, Cylindrical Direct Numerical Simulation code CYLDNS and molecular dynamics code for simulating radiation damages in diamond crystals DGR are described. In the vectorization part, the vectorization of two and three dimensional discrete ordinates simulation code DORT-TORT, gas dynamics analysis code FLOWGR and relativistic Boltzmann-Uehling-Uhlenbeck simulation code RBUU are described. And then, in the porting part, the porting of reactor safety analysis code RELAP5/MOD3.2 and RELAP5/MOD3.2.1.2, nuclear data processing system NJOY and 2-D multigroup discrete ordinate transport code TWOTRAN-II are described. And also, a survey for the porting of command-driven interactive data analysis plotting program IPLOT are described. (author)

  7. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (vectorization). Progress report fiscal 1996

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki; Kawai, Wataru [Fujitsu Ltd., Tokyo (Japan); Kawasaki, Nobuo [and others

    1997-12-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the vectorization. In this vectorization part, the vectorization of two and three dimensional discrete ordinates simulation code DORT-TORT, gas dynamics analysis code FLOWGR and relativistic Boltzmann-Uehling-Uhlenbeck simulation code RBUU are described. In the parallelization part, the parallelization of 2-Dimensional relativistic electromagnetic particle code EM2D, Cylindrical Direct Numerical Simulation code CYLDNS and molecular dynamics code for simulating radiation damages in diamond crystals DGR are described. And then, in the porting part, the porting of reactor safety analysis code RELAP5/MOD3.2 and RELAP5/MOD3.2.1.2, nuclear data processing system NJOY and 2-D multigroup discrete ordinate transport code TWOTRAN-II are described. And also, a survey for the porting of command-driven interactive data analysis plotting program IPLOT are described. (author)

  8. An approach to validation of coupled CFD and system thermal-hydraulics codes

    International Nuclear Information System (INIS)

    This paper discusses the development of approach and experimental facility for the validation of coupled Computational Fluid Dynamics (CFD) and System Thermal Hydraulics (STH) codes. The validation of a coupled code requires experiments which feature two way feedback between the component (CFD sub-domain) and the system (STH sub-domain). We present results of CFD analysis that are used in the development of a flexible design for the TALL-3D experimental facility. The facility consists of a lead-bismuth thermal-hydraulic loop operating in forced and natural circulation regimes with a heated pool-type 3D test section. The goal of the design is to achieve a feedback between mixing and stratification phenomena in the 3D tests section and forced / natural circulation flow conditions in the loop. Finally, we discuss the development of an experimental validation matrix for validation of coupled STH and CFD codes that considers the key physical phenomena of interest. (author)

  9. Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

    International Nuclear Information System (INIS)

    The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

  10. Analysis Code for High Gradient Dielectric Insulator Surface Breakdown

    Energy Technology Data Exchange (ETDEWEB)

    Ives, Robert Lawrence [Calabazas Creek Research, Inc.; Verboncoeur, John [University of California - Berkeley; Aldan, Manuel [University of California, Berkeley

    2010-05-30

    High voltage (HV) insulators are critical components in high-energy, accelerator and pulsed power systems that drive diverse applications in the national security, nuclear weapons science, defense and industrial arenas. In these systems, the insulator may separate vacuum/non-vacuum regions or conductors with high electrical field gradients. These insulators will often fail at electric fields over an order of magnitude lower than their intrinsic dielectric strength due to flashover at the dielectric interface. Decades of studies have produced a wealth of information on fundamental processes and mechanisms important for flashover initiation, but only for relatively simple insulator configurations in controlled environments. Accelerator and pulsed power system designers are faced with applying the fundamental knowledge to complex, operational devices with escalating HV requirements. Designers are forced to rely on “best practices” and expensive prototype testing, providing boundaries for successful operation. However, the safety margin is difficult to estimate, and system design must be very conservative for situations where testing is not practicable, or replacement of failed parts is disruptive or expensive. The Phase I program demonstrated the feasibility of developing an advanced code for modeling insulator breakdown. Such a code would be of great interest for a number of applications, including high energy physics, microwave source development, fusion sciences, and other research and industrial applications using high voltage devices.

  11. Assessment of RELAP5/CANDU+ code for regulatory auditing analysis of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Kim, Hho Jung; Yang, Chae Yong

    2001-12-15

    The objectives of this study are to undertake the verification and validation of RELAP5/CANDU+ code, which is developed in this project, by simulating the B8711 test of RD-14 facility, and to examine the properties of this code by doing the sensitivity analysis for experimental prediction modes about thermal-hydraulics phenomena in CANDU reactor systems added to this code. The B8711 test was an experiment of a 45% ROH break for simulating large LOCA. Also, in this study, the methods for making input cards related to CANDU options are described, so that some users can use the RELAP5/CANDU+ code with easy. RELAP/CANDU+ code can choose the options of Henry-Fauske mode, Ransom-Trapp model, and Moody model for prediction of the critical mass flow. It is examined that Henry-Fauske model and Ransom-Trapp model are considered properly, but Moody model is still required to be improved. Heat transfer correlations available in RELAP5/CANDU+ code for CANDU-type reactors are a horizontal stratified model, a fuel heat-up model and D2O/H2O CHF correlations, and these models take an important role to improve the predictability of the experimental procedures. It is concluded that RELAP5/CANDU+ code is useful for the auditing of the accident analysis of CANDU reactors, and the results of the sensitivity analysis for thermal-hydraulic models examined in this study are valuable for the actual auditing of real CANDU-type power plants.

  12. MULTI-KENO: a Monte Carlo code for criticality safety analysis

    International Nuclear Information System (INIS)

    Modifying the Monte Carlo code KENO-IV, the MULTI-KENO code was developed for criticality safety analysis. The following functions were added to the code; (1) to divide a system into many sub-systems named super boxes where the size of box types in each super box can be selected independently, (2) to output graphical view of a system for examining geometrical input data, (3) to solve fixed source problems, (4) to permit intersection of core boundaries and inner geometries, (5) to output ANISN type neutron balance table. With the above function (1), many cases which had to be applied a general geometry option of KENO-IV, became to be treated as box type geometry. In such a case, input data became simpler and required computer time became shorter than those of KENO-IV. This code is now available for the FACOM-M200 computer and the CDC 6600 computer. This report is a computer code manual for MULTI-KENO. (author)

  13. KKS编码系统在电厂设备基础信息库的应用分析%Application Analysis of KKS Coding System in Equipment Foundation Database of Power Plants

    Institute of Scientific and Technical Information of China (English)

    周惠平

    2014-01-01

    According to the KKS coding system completed during power plant design,a novel mode of equipment management combined KKS coding system and equipment foundation database after power plant construction completed was introduce in this study. Also, the characteristics of equipment foundation database with KKS coding system were analyzed. The conclusion is that the unified equipment coding system of power plants could realize automation,informatization and information sharing of equipment management of power plants.%针对在电厂工程设计阶段完成的KKS编码系统,介绍了电厂建成后将KKS编码系统实现扩展,并与电厂设备基础信息库结合的新的设备管理模式。分析了带KKS编码系统的电厂设备基础信息库的特点,结论是基于统一的电厂设备编码体系,可实现电厂设备管理的自动化、信息化、信息共享化。

  14. Systemizers Are Better Code-Breakers: Self-Reported Systemizing Predicts Code-Breaking Performance in Expert Hackers and Naïve Participants

    Science.gov (United States)

    Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena

    2016-01-01

    Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders—attention to detail and systemizing—may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings. PMID:27242491

  15. New Parallel Interference Cancellation for Convolutionally Coded CDMA Systems

    Institute of Scientific and Technical Information of China (English)

    Xu Guo-xiong; Gan Liang-cai; Huang Tian-xi

    2004-01-01

    Based on BCJR algorithm proposed by Bahl et al and linear soft decision feedback, a reduced-complexity parallel interference cancellation (simplified PIC) for convolutionally coded DS CDMA systems is proposed. By computer simulation, we compare the simplified PIC with the exact PIC. It shows that the simplified PIC can achieve the performance close to the exact PIC if the mean values of coded symbols are linearly computed in terms of the sum of initial a prior log-likelihood rate (LLR) and updated a prior LLR, while a significant performance loss will occur if the mean values of coded symbols are linearly computed in terms of the updated a prior LLR only. Meanwhile, we also compare the simplified PIC with MF receiver and conventional PICs. The simulation results show that the simplified PIC dominantly outperforms the MF receiver and conventional PICs, at signal-noise rate (SNR) of 7 dB, for example, the bit error rate is about 10-4 for the simplified PIC, which is far below that of matched-filter receiver and conventional PIC.

  16. System Design Considerations In Bar-Code Laser Scanning

    Science.gov (United States)

    Barkan, Eric; Swartz, Jerome

    1984-08-01

    The unified transfer function approach to the design of laser barcode scanner signal acquisition hardware is considered. The treatment of seemingly disparate system areas such as the optical train, the scanning spot, the electrical filter circuits, the effects of noise, and printing errors is presented using linear systems theory. Such important issues as determination of depth of modulation, filter specification, tolerancing of optical components, and optimi-zation of system performance in the presence of noise are discussed. The concept of effective spot size to allow for impact of optical system and analog processing circuitry upon depth of modulation is introduced. Considerations are limited primarily to Gaussian spot profiles, but also apply to more general cases. Attention is paid to realistic bar-code symbol models and to implications with respect to printing tolerances.

  17. Multiple Description Coding for Closed Loop Systems over Erasure Channels

    DEFF Research Database (Denmark)

    Østergaard, Jan; Quevedo, Daniel

    2013-01-01

    ) and the decoder (plant). The feedback channel from the decoder to the encoder is assumed noiseless. Since the forward channel is digital, we need to employ quantization.We combine two techniques to enhance the reliability of the system. First, in order to guarantee that the system remains stable during packet......In this paper, we consider robust source coding in closed-loop systems. In particular, we consider a (possibly) unstable LTI system, which is to be stabilized via a network. The network has random delays and erasures on the data-rate limited (digital) forward channel between the encoder (controller....... In particular, we transmit M redundant packets, which are constructed such that when receiving any J packets, the current control signal as well as J-1 future control signals can be reliably reconstructed at the decoder. We prove stability subject to quantization constraints, random dropouts, and delays...

  18. Verification of structural analysis computer codes in nuclear engineering

    International Nuclear Information System (INIS)

    Sources of potential errors, which can take place during use of finite element method based computer programs, are described in the paper. The magnitude of errors was defined as acceptance criteria for those programs. Error sources are described as they are treated by 'National Agency for Finite Element Methods and Standards (NAFEMS)'. Specific verification examples are used from literature of Nuclear Regulatory Commission (NRC). Example of verification is made on PAFEC-FE computer code for seismic response analyses of piping systems by response spectrum method. (author)

  19. Obfuscated Malicious Code Detection with Path Condition Analysis

    OpenAIRE

    Wenqing Fan; Xue Lei; Jing An

    2014-01-01

    Code obfuscation is one of the main methods to hide malicious code. This paper proposes a new dynamic method which can effectively detect obfuscated malicious code. This method uses ISR to conduct dynamic debugging. The constraint solving during debugging process can detect deeply hidden malicious code by covering different execution paths. Besides, for malicious code that reads external resources, usually the detection of abnormal behaviors can only be detected by taking the resources into c...

  20. Nonterminals, homomorphisms and codings in different variations of OL-systems. II. Nondeterministic systems

    DEFF Research Database (Denmark)

    Nielsen, Mogens; Rozenberg, Grzegorz; Salomaa, Arto;

    1974-01-01

    Continuing the work begun in Part I of this paper, we consider now variations of nondeterministic OL-systems. The present Part II of the paper contains a systematic classification of the effect of nonterminals, codings, weak codings, nonerasing homomorphisms and homomorphisms for all basic variat...

  1. CFC (Comment-First-Coding)--A Simple yet Effective Method for Teaching Programming to Information Systems Students

    Science.gov (United States)

    Sengupta, Arijit

    2009-01-01

    Programming courses have always been a difficult part of an Information Systems curriculum. While we do not train Information Systems students to be developers, understanding how to build a system always gives students an added perspective to improve their system design and analysis skills. This teaching tip presents CFC (Comment-First-Coding)--a…

  2. Evaluation of Recent Upgrades to the NESS (Nuclear Engine System Simulation) Code

    Science.gov (United States)

    Fittje, James E.; Schnitzler, Bruce G.

    2008-01-01

    The Nuclear Thermal Rocket (NTR) concept is being evaluated as a potential propulsion technology for exploratory expeditions to the moon, Mars, and beyond. The need for exceptional propulsion system performance in these missions has been documented in numerous studies, and was the primary focus of a considerable effort undertaken during the Rover/NERVA program from 1955 to 1973. The NASA Glenn Research Center is leveraging this past NTR investment in their vehicle concepts and mission analysis studies with the aid of the Nuclear Engine System Simulation (NESS) code. This paper presents the additional capabilities and upgrades made to this code in order to perform higher fidelity NTR propulsion system analysis and design, and a comparison of its results to the Small Nuclear Rocket Engine (SNRE) design.

  3. Electronic health record standards, coding systems, frameworks, and infrastructures

    CERN Document Server

    Sinha, Pradeep K; Bendale, Prashant; Mantri, Manisha; Dande, Atreya

    2013-01-01

    Discover How Electronic Health Records Are Built to Drive the Next Generation of Healthcare Delivery The increased role of IT in the healthcare sector has led to the coining of a new phrase ""health informatics,"" which deals with the use of IT for better healthcare services. Health informatics applications often involve maintaining the health records of individuals, in digital form, which is referred to as an Electronic Health Record (EHR). Building and implementing an EHR infrastructure requires an understanding of healthcare standards, coding systems, and frameworks. This book provides an

  4. HYDRA-II: A hydrothermal analysis computer code: Volume 3, Verification/validation assessments

    International Nuclear Information System (INIS)

    HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum are enhanced by the incorporation of directional porosities and permeabilities that aid in modeling solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated procedures are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume I - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. Volume II - User's Manual contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a model problem. This volume, Volume III - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. This volume also documents comparisons between the results of simulations of single- and multiassembly storage systems and actual experimental data. 11 refs., 55 figs., 13 tabs

  5. Renewable Energy Requirements for Future Building Codes: Energy Generation and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Russo, Bryan J.; Weimar, Mark R.; Dillon, Heather E.

    2011-09-30

    As the model energy codes are improved to reach efficiency levels 50 percent greater than current codes, installation of on-site renewable energy generation is likely to become a code requirement. This requirement will be needed because traditional mechanisms for code improvement, including the building envelope, mechanical systems, and lighting, have been maximized at the most cost-effective limit.

  6. [Data coding in the Israeli healthcare system - do choices provide the answers to our system's needs?].

    Science.gov (United States)

    Zelingher, Julian; Ash, Nachman

    2013-05-01

    The IsraeLi healthcare system has undergone major processes for the adoption of health information technologies (HIT), and enjoys high Levels of utilization in hospital and ambulatory care. Coding is an essential infrastructure component of HIT, and ts purpose is to represent data in a simplified and common format, enhancing its manipulation by digital systems. Proper coding of data enables efficient identification, storage, retrieval and communication of data. UtiLization of uniform coding systems by different organizations enables data interoperability between them, facilitating communication and integrating data elements originating in different information systems from various organizations. Current needs in Israel for heaLth data coding include recording and reporting of diagnoses for hospitalized patients, outpatients and visitors of the Emergency Department, coding of procedures and operations, coding of pathology findings, reporting of discharge diagnoses and causes of death, billing codes, organizational data warehouses and national registries. New national projects for cLinicaL data integration, obligatory reporting of quality indicators and new Ministry of Health (MOH) requirements for HIT necessitate a high Level of interoperability that can be achieved only through the adoption of uniform coding. Additional pressures were introduced by the USA decision to stop the maintenance of the ICD-9-CM codes that are also used by Israeli healthcare, and the adoption of ICD-10-C and ICD-10-PCS as the main coding system for billing purpose. The USA has also mandated utilization of SNOMED-CT as the coding terminology for the ELectronic Health Record problem list, and for reporting quality indicators to the CMS. Hence, the Israeli MOH has recently decided that discharge diagnoses will be reported using ICD-10-CM codes, and SNOMED-CT will be used to code the cLinical information in the EHR. We reviewed the characteristics, strengths and weaknesses of these two coding

  7. Coded aper ture compressive imaging array applied for surveillance systems

    Institute of Scientific and Technical Information of China (English)

    Jing Chen; Yongtian Wang; Hanxiao Wu

    2013-01-01

    This paper proposes an application of compressive imaging systems to the problem of wide-area video surveil ance systems. A paral el coded aperture compressive imaging sys-tem and a corresponding motion target detection algorithm in video using compressive image data are developed. Coded masks with random Gaussian, Toeplitz and random binary are utilized to simulate the compressive image respectively. For compres-sive images, a mixture of the Gaussian distribution is applied to the compressed image field to model the background. A simple threshold test in compressive sampling image is used to declare motion objects. Foreground image retrieval from underdetermined measurement using the total variance optimization algorithm is explored. The signal-to-noise ratio (SNR) is employed to evalu-ate the image quality recovered from the compressive sampling signals, and receiver operation characteristic (ROC) curves are used to quantify the performance of the motion detection algo-rithm. Experimental results demonstrate that the low dimensional compressed imaging representation is sufficient to determine spa-tial motion targets. Compared with the random Gaussian and Toeplitz mask, motion detection algorithms using the random bi-nary phase mask can yield better detection results. However using the random Gaussian and Toeplitz phase mask can achieve high resolution reconstructed images.

  8. The Application Programming Interface for the PVMEXEC Program and Associated Code Coupling System

    Energy Technology Data Exchange (ETDEWEB)

    Walter L. Weaver III

    2005-03-01

    This report describes the Application Programming Interface for the PVMEXEC program and the code coupling systems that it implements. The information in the report is intended for programmers wanting to add a new code into the coupling system.

  9. Optimization and Validation of the Developed Uranium Isotopic Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    γ-ray spectroscopy is a representative non-destructive assay for nuclear material, and less time-consuming and less expensive than the destructive analysis method. The destructive technique is more precise than NDA technique, however, there is some correction algorithm which can improve the performance of γ-spectroscopy. For this reason, an analysis code for uranium isotopic analysis is developed by Applied Nuclear Physics Group in Seoul National University. Overlapped γ- and x-ray peaks in the 89-101 keV X{sub α}-region are fitted with Gaussian and Lorentzian distribution peak functions, tail and background functions. In this study, optimizations for the full-energy peak efficiency calibration and fitting parameters of peak tail and background are performed, and validated with 24 hour acquisition of CRM uranium samples. The optimization of peak tail and background parameters are performed with the validation by using CRM uranium samples. The analysis performance is improved in HEU samples, but more optimization of fitting parameters is required in LEU sample analysis. In the future, the optimization research about the fitting parameters with various type of uranium samples will be performed. {sup 234}U isotopic analysis algorithms and correction algorithms (coincidence effect, self-attenuation effect) will be developed.

  10. Digital Image Analysis for DETCHIP(®) Code Determination.

    Science.gov (United States)

    Lyon, Marcus; Wilson, Mark V; Rouhier, Kerry A; Symonsbergen, David J; Bastola, Kiran; Thapa, Ishwor; Holmes, Andrea E; Sikich, Sharmin M; Jackson, Abby

    2012-08-01

    DETECHIP(®) is a molecular sensing array used for identification of a large variety of substances. Previous methodology for the analysis of DETECHIP(®) used human vision to distinguish color changes induced by the presence of the analyte of interest. This paper describes several analysis techniques using digital images of DETECHIP(®). Both a digital camera and flatbed desktop photo scanner were used to obtain Jpeg images. Color information within these digital images was obtained through the measurement of red-green-blue (RGB) values using software such as GIMP, Photoshop and ImageJ. Several different techniques were used to evaluate these color changes. It was determined that the flatbed scanner produced in the clearest and more reproducible images. Furthermore, codes obtained using a macro written for use within ImageJ showed improved consistency versus pervious methods. PMID:25267940

  11. Calculation of the Novovoronezh Recriticality Experiment with the KARATE-440 code system

    Energy Technology Data Exchange (ETDEWEB)

    Hegyi, György, E-mail: ghegyi@aeki.kfki.hu [MTA KFKI Atomic Energy Research Institute, Budapest (Hungary)

    2011-07-01

    In this paper the results of KARATE-440 calculations on Novovoronezh NPP Recriticality Experiment are presented, the corresponding parameters are analyzed. The simulation of the processes and the comparison of the results with the measurements are of particular interest as these efforts make our code to be validated in a higher level. The KARATE-440 code system has been developed and applied for VVER-440 core analysis during near twenty years, as a close collaboration among the developers and the specialists at the 4 Hungarian nuclear power units. KARATE is now a mature, demonstrated, complete and integrated system of computer codes and procedures that provide full and independent VVER core analysis capabilities. Even if only some well defined states of the experiment were simulated, satisfactory agreement was found between measured and calculated data. The results present evidence that the KARATE- 440 code package can adequately model the reactor states in a wide range of performance parameters and the special core type referred in the experiment so it is acceptable for neutronic analysis of all the VVER-440 NPP's. (author)

  12. LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System

    International Nuclear Information System (INIS)

    1 - Description of program or function: The LOLA-SYSTEM is a part of the JEN-UPM code package for PWR fuel management, scope or design calculations. It is a code package for core burnup calculations using nodal theory based on a FLARE type code. The LOLA-SYSTEM includes four modules: the first one (MELON-3) generates the constants of the K-inf and M2 correlations to be input into SIMULA-3. It needs the K-inf and M2 fuel assembly values at different conditions of moderator temperature, Boron concentration, burnup, etc., which are provided by MARIA fuel assembly calculations. The main module (SIMULA-3) is the core burnup calculation code in three dimensions and one group of energy. It normally uses a geometrical representation of one node per fuel assembly or per quarter of fuel assembly. It has included a thermal hydraulic feedback on flow and voids and criticality searches on boron concentration and control rods insertion. The CONCON code makes the calculation of the albedo, transport factors, K-inf and M2 correction factors to be input into SIMULA-3. The calculation is made in the XY transversal plane. The CONAXI code is similar to CONCON, but in the axial direction. 2 - Method of solution: MELON-3 makes a mean squares fit of K-inf and M2 values at different conditions in order to determine the constants of the feedback correlations. SIMULA-3 uses a modified one-group nodal theory, with a new transport kernel that provides the same node interface leakages as a fine mesh diffusion calculation. CONCON and CONAXI determine the transport and correction factors, as well as the albedo, to be input into SIMULA-3. They are determined by a method of leakages equivalent to the detailed diffusion calculation of CARMEN or VENTURE; these factors also include the heterogeneity effects inside the node. 3 - Restrictions on the complexity of the problem: Number of axial nodes less than or equal 34. Number of material types less than or equal 30. Number of fuel assembly types less

  13. Thermal-hydraulic analysis for reactor vessel upper-head small break LOCA using SPACE code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [Korea Hydro and Nuclear Power Co., Central Research Inst., Daejeon (Korea, Republic of)

    2015-08-15

    A small break loss of coolant accident (SBLOCA) in upper-head of a reactor vessel at OPR1000 was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. To assess the capability of SPACE code, upper-head SBLOCA with full plant safeguards was simulated, and compared with results of MARS-KS code. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. Based on the observed thermal-hydraulic features, simulations with the failure of partial plant safeguards were conducted to analyze the safety and performance of OPR1000. Effects of failure to scram and high-pressure safety injection (HPSI) were investigated, and safety assessment was evaluated according to operator actions. Comparative study without any emergency core cooling systems (ECCS) was also conducted to judge the severity of the break location. From the results, this indicated that SPACE code has capabilities to simulate upper-head SBLOCA, and OPR1000 was evaluated to have sufficient safety margin with the application of proper emergency operating procedures.

  14. Preliminary core mechanics analysis for KALIMER by CRAMP code

    International Nuclear Information System (INIS)

    CRAMP code is designed to solve the problem of mutually interacting and distorting sub-assemblies in a fast breeder reactor. It is the UK's main core mechanics design tool and is currently being used in the design of EFR. This report contains the results of preliminary core mechanics calculations for KALIMER core configuration by the updated version of CRAMP code. The base case calculation s on KALIMER core, and the sensitivity studies (to investigate effect of main design parameter) are carried out by the code which was updated with material subroutine in CRAMP to model the characteristics of HT9. Sensitivity studies include following cases; (1) with gaps at LRP and URP reduced to 0.4 mm at 386 dg C (2) with 0.2 mm radial clearance around both nosepiece at seals (3) with flexibility at LRP reduced by a factor of 2 (4) with stiffness of nosepiece increased by a factor of 2 (5) with reduced creep (6) combined with gap reduced 0.4 mm and 0.2 mm clearance at seals (7) with IVS position replaced to dummy ducts (8) with initial bow at every duct except S/A 1. From each calculation, the data obtained and compared are as follows; (a) contact forces between pads (b) gaps between pads (c) duct dynamic behavior of duct bowing and dilation i.e. the variation of bowing and dilation with time) (d) reactivity change (e) maximum LRP contact force, maximum URP contact force and maximum nosepiece force vs interval number for the base case. The design requirements and the specifications for KALIMER assembly ducts are reviewed, and preliminary core mechanics analysis for KALIMER core configuration are carried out. (Author). 7 refs., 2 tabs., 50 figs

  15. HYDRA-II: A hydrothermal analysis computer code: Volume 2, User's manual

    International Nuclear Information System (INIS)

    HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite-difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum incorporate directional porosities and permeabilities that are available to model solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated methods are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume 1 - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. This volume, Volume 2 - User's Manual, contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a sample problem. The final volume, Volume 3 - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. 6 refs

  16. HYDRA-II: A hydrothermal analysis computer code: Volume 2, User's manual

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.; Lowery, P.S.; Lessor, D.L.

    1987-09-01

    HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite-difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum incorporate directional porosities and permeabilities that are available to model solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated methods are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume 1 - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. This volume, Volume 2 - User's Manual, contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a sample problem. The final volume, Volume 3 - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. 6 refs.

  17. PWR experimental benchmark analysis using WIMSD and PRIDE codes

    International Nuclear Information System (INIS)

    Highlights: • PWR experimental benchmark calculations were performed using WIMSD and PRIDE codes. • Various models for lattice cell homogenization were used. • Multiplication factors, power distribution and reaction rates were studied. • The effect of cross section libraries on these parameters was analyzed. • The results were compared with experimental and reported results. - Abstract: The PWR experimental benchmark problem defined by ANS was analyzed using WIMSD and PRIDE codes. Different modeling methodologies were used to calculate the infinite and effective multiplication factors. Relative pin power distributions were calculated for infinite lattice and critical core configurations, while reaction ratios were calculated for infinite lattice only. The discrete ordinate method (DSN) and collision probability method (PERSEUS) were used in each calculation. Different WIMSD cross-section libraries based on ENDF/B-VI.8, ENDF/B-VII.0, IAEA, JEF-2.2, JEFF-3.1 and JENDL-3.2 nuclear data files were also employed in the analyses. Comparison was made with experimental data and other reported results in order to find a suitable strategy for PWR analysis

  18. Axisymmetric Plume Simulations with NASA's DSMC Analysis Code

    Science.gov (United States)

    Stewart, B. D.; Lumpkin, F. E., III

    2012-01-01

    A comparison of axisymmetric Direct Simulation Monte Carlo (DSMC) Analysis Code (DAC) results to analytic and Computational Fluid Dynamics (CFD) solutions in the near continuum regime and to 3D DAC solutions in the rarefied regime for expansion plumes into a vacuum is performed to investigate the validity of the newest DAC axisymmetric implementation. This new implementation, based on the standard DSMC axisymmetric approach where the representative molecules are allowed to move in all three dimensions but are rotated back to the plane of symmetry by the end of the move step, has been fully integrated into the 3D-based DAC code and therefore retains all of DAC s features, such as being able to compute flow over complex geometries and to model chemistry. Axisymmetric DAC results for a spherically symmetric isentropic expansion are in very good agreement with a source flow analytic solution in the continuum regime and show departure from equilibrium downstream of the estimated breakdown location. Axisymmetric density contours also compare favorably against CFD results for the R1E thruster while temperature contours depart from equilibrium very rapidly away from the estimated breakdown surface. Finally, axisymmetric and 3D DAC results are in very good agreement over the entire plume region and, as expected, this new axisymmetric implementation shows a significant reduction in computer resources required to achieve accurate simulations for this problem over the 3D simulations.

  19. Conjugate heat transfer analysis using the Calore and Fuego codes.

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Nicholas Donald, Jr.

    2007-09-01

    Full coupling of the Calore and Fuego codes has been exercised in this report. This is done to allow solution of general conjugate heat transfer applications that require more than a fluid flow analysis with a very simple conduction region (solved using Fuego alone) or more than a complex conduction/radiation analysis using a simple Newton's law of cooling boundary condition (solved using Calore alone). Code coupling allows for solution of both complex fluid and solid regions, with or without thermal radiation, either participating or non-participating. A coupled physics model is developed to compare to data taken from a horizontal concentric cylinder arrangement using the Penlight heating apparatus located at the thermal test complex (TTC) at Sandia National Laboratories. The experimental set-up requires use of a conjugate heat transfer analysis including conduction, nonparticipating thermal radiation, and internal natural convection. The fluids domain in the model is complex and can be characterized by stagnant fluid regions, laminar circulation, a transition regime, and low-level turbulent regions, all in the same domain. Subsequently, the fluids region requires a refined mesh near the wall so that numerical resolution is achieved. Near the wall, buoyancy exhibits its strongest influence on turbulence (i.e., where turbulence conditions exist). Because low-Reynolds number effects are important in anisotropic natural convective flows of this type, the {ovr {nu}{sup 2}}-f turbulence model in Fuego is selected and compared to results of laminar flow only. Coupled code predictions are compared to temperature measurements made both in the solid regions and a fluid region. Turbulent and laminar flow predictions are nearly identical for both regions. Predicted temperatures in the solid regions compare well to data. The largest discrepancies occur at the bottom of the annulus. Predicted temperatures in the fluid region, for the most part, compare well to data. As

  20. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules, F9-F11

    International Nuclear Information System (INIS)

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with three of the functional modules in the code. Those are the Morse-SGC for the SCALE system, Heating 7.2, and KENO V.a. The manual describes the latest released versions of the codes

  1. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules, F9-F11

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with three of the functional modules in the code. Those are the Morse-SGC for the SCALE system, Heating 7.2, and KENO V.a. The manual describes the latest released versions of the codes.

  2. ASFRE: a computer code for single-phase subchannel thermal hydraulic analysis of LMFBR single subassembly

    International Nuclear Information System (INIS)

    The objectives of this work is to develop a computer code ASFRE which analyzes 3D-thermo-hydraulic behaviors of coolant and fuel pins in an LMFBR subassembly under accident conditions such as the local blockage, loss of flow and transient over power accident conditions. Analytical models, calculation procedures and sample calculations for typical experiments are described. The ASFRE code consists of two parts, namely coolant calculation part and fuel pin calculation. The coolant thermal-hydraulic analysis employs basically subchannel analysis approach and the program solves transient mass, momentum and energy conservation equations. The fuel pin thermal analysis program solves transient heat conduction equations by finite difference method in cylindrical coordinate system. Fuel temperature distribution and thermal expansion are calculated taking into account of intra/inter-pin-flux-depression and fuel restructuring. And wire wrap spacer effects for coolant behavior and heat loss through the wrapper tube are also simulated. (author)

  3. Overview of particle and heavy ion transport code system PHITS

    International Nuclear Information System (INIS)

    Highlights: • We developed a general-purpose Monte Carlo particle transport code PHITS. • PHITS can deal with the transport of nearly all particles over wide energy ranges. • More than 1500 researchers have been used PHITS for various applications. • Physics models and special functions implemented in PHITS are briefly summarized. - Abstract: A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research Organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development’s Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1500 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-04-01

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

  5. Source Code Analysis to Remove Security Vulnerabilities in Java Socket Programs: A Case Study

    Directory of Open Access Journals (Sweden)

    Natarajan Meghanathan

    2013-02-01

    Full Text Available This paper presents the source code analysis of a file reader server socket program (connection-orientedsockets developed in Java, to illustrate the identification, impact analysis and solutions to remove fiveimportant software security vulnerabilities, which if left unattended could severely impact the serverrunning the software and also the network hosting the server. The five vulnerabilities we study in thispaper are: (1 Resource Injection, (2 Path Manipulation, (3 System Information Leak, (4 Denial ofService and (5 Unreleased Resource vulnerabilities. We analyze the reason why each of thesevulnerabilities occur in the file reader server socket program, discuss the impact of leaving themunattended in the program, and propose solutions to remove each of these vulnerabilities from theprogram. We also analyze any potential performance tradeoffs (such as increase in code size and loss offeatures that could arise while incorporating the proposed solutions on the server program. Theproposed solutions are very generic in nature, and can be suitably modified to correct any suchvulnerabilities in software developed in any other programming language. We use the Fortify SourceCode Analyzer to conduct the source code analysis of the file reader server program, implemented on aWindows XP virtual machine with the standard J2SE v.7 development kit.

  6. Knitting music and programming: Reflections on the frontiers of source code analysis

    OpenAIRE

    Gold, N.

    2011-01-01

    Source Code Analysis and Manipulation (SCAM) underpins virtually every operational software system. Despite the impact and ubiquity of SCAM principles and techniques in software engineering, there are still frontiers to be explored. Looking "inward" to existing techniques, one finds frontiers of performance, efficiency, accuracy, and usability, looking "outward" one finds new languages, new problems, and thus new approaches. This paper presents a reflective framework for characterizing source...

  7. FPGA based digital phase-coding quantum key distribution system

    Science.gov (United States)

    Lu, XiaoMing; Zhang, LiJun; Wang, YongGang; Chen, Wei; Huang, DaJun; Li, Deng; Wang, Shuang; He, DeYong; Yin, ZhenQiang; Zhou, Yu; Hui, Cong; Han, ZhengFu

    2015-12-01

    Quantum key distribution (QKD) is a technology with the potential capability to achieve information-theoretic security. Phasecoding is an important approach to develop practical QKD systems in fiber channel. In order to improve the phase-coding modulation rate, we proposed a new digital-modulation method in this paper and constructed a compact and robust prototype of QKD system using currently available components in our lab to demonstrate the effectiveness of the method. The system was deployed in laboratory environment over a 50 km fiber and continuously operated during 87 h without manual interaction. The quantum bit error rate (QBER) of the system was stable with an average value of 3.22% and the secure key generation rate is 8.91 kbps. Although the modulation rate of the photon in the demo system was only 200 MHz, which was limited by the Faraday-Michelson interferometer (FMI) structure, the proposed method and the field programmable gate array (FPGA) based electronics scheme have a great potential for high speed QKD systems with Giga-bits/second modulation rate.

  8. Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code

    OpenAIRE

    C. A. M. Silva; J. A. D. Salomé; B. T. Guerra; Pereira, C; Costa, A. L.; Veloso, M. A. F.; M. A. B. C. Menezes; Dalle, H. M.

    2014-01-01

    In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport) code. The sensitivity analyses included sma...

  9. Motion-compensated coding and frame rate up-conversion: models and analysis.

    Science.gov (United States)

    Dar, Yehuda; Bruckstein, Alfred M

    2015-07-01

    Block-based motion estimation (ME) and motion compensation (MC) techniques are widely used in modern video processing algorithms and compression systems. The great variety of video applications and devices results in diverse compression specifications, such as frame rates and bit rates. In this paper, we study the effect of frame rate and compression bit rate on block-based ME and MC as commonly utilized in inter-frame coding and frame rate up-conversion (FRUC). This joint examination yields a theoretical foundation for comparing MC procedures in coding and FRUC. First, the video signal is locally modeled as a noisy translational motion of an image. Then, we theoretically model the motion-compensated prediction of available and absent frames as in coding and FRUC applications, respectively. The theoretic MC-prediction error is studied further and its autocorrelation function is calculated, yielding useful separable-simplifications for the coding application. We argue that a linear relation exists between the variance of the MC-prediction error and temporal distance. While the relevant distance in MC coding is between the predicted and reference frames, MC-FRUC is affected by the distance between the frames available for interpolation. We compare our estimates with experimental results and show that the theory explains qualitatively the empirical behavior. Then, we use the models proposed to analyze a system for improving of video coding at low bit rates, using a spatio-temporal scaling. Although this concept is practically employed in various forms, so far it lacked a theoretical justification. We here harness the proposed MC models and present a comprehensive analysis of the system, to qualitatively predict the experimental results.

  10. Development of methods for the analysis of accident scenarios with steam line breaks and boron dilution by the help of the code system ATHLET-DYN3D. Final report. Pt. 1

    International Nuclear Information System (INIS)

    Libraries of two-group neutron-diffusion parameters for a Siemens-KWU-Konvoi Pressurized Water Reactor have been generated at Forschungszentrum Rossendorf and TUeV Bau und Betrieb GmbH by using the codes HELIOS and CASMO, respectively. The libraries have been coupled to the reactor-dynamics code DYN3D. For a generic PWR core containing MOX fuel elements, DYN3D macro-burnup calculations and the calculation of different operation states have been carried out. The results will be used for the investigation of possible accident scenarios. Reactivity coefficients calculated by DYN3D are needed for accident analyses by the 1-D thermal-hydraulic code ATHLET. Using the cross section data, more detailed analyses can be carried out by applying the coupled-code system DYN3D-ATHLET, considering 3D neutron kinetics. The comparison of the results calculated by DYN3D with two different diffusion-parameter libraries can give an idea of how uncertainties in diffusion data influence the accuracy of reactor simulation. (orig.)

  11. A video coding system for sign language communication at low bit rates

    OpenAIRE

    Agrafiotis, D.; Canagarajah, CN; Bull, DR; Kyle, J; Seers, H; Dye, M

    2004-01-01

    The ability to communicate remotely through the use of video as promised by wireless networks and already practised over fixed networks, is for deaf people as important as voice telephony is for hearing people. Sign languages are visual-spatial languages and as such demand good image quality for interaction and understanding. In this paper, based on analysis of the viewer's perceptual behavior and the video content involved we propose a sign language video coding system using foveated process...

  12. MACRO1: a code to test a methodology for analyzing nuclear-waste management systems

    International Nuclear Information System (INIS)

    The code is primarily a manager of probabilistic data and deterministic mathematical models. The user determines the desired aggregation of the available models into a composite model of a physical system. MACRO1 then propagates the finite probability distributions of the inputs to the model to finite probability distributions over the outputs. MACRO1 has been applied to a sample analysis of a nuclear-waste repository, and its results compared satisfactorily with previously obtained Monte Carlo statistics

  13. Two Schemes of Blind MMSE Multiuser Receiver for Space-Time Coded CDMA Systems

    Institute of Scientific and Technical Information of China (English)

    LU Min; XU Chang-jiang; FENG Guang-zeng

    2004-01-01

    Minimum Mean Square Error (MMSE) multiuser detection yields the highest output SINR among all linear detectors. The blind MMSE linear detector can be implemented with batch processes or sequential processes. In this paper, according to the different implementations of blind detectors, the authors analyze two schemes of the blind MMSE multiuser receiver for space-time coded CDMA Systems and make a comparison between both schemes by the theoretical analysis and numerical simulations.

  14. A design of a wavelength-hopping time-spreading incoherent optical code division multiple access system

    International Nuclear Information System (INIS)

    We present the architecture and code design for a highly scalable, 2.5 Gb/s per user optical code division multiple access (OCDMA) system. The system is scalable to 100 potential and more than 10 simultaneous users, each with a bit error rate (BER) of less than 10-9. The system architecture uses a fast wavelength-hopping, time-spreading codes. Unlike frequency and phase sensitive coherent OCDMA systems, this architecture utilizes standard on off keyed optical pulses allocated in the time and wavelength dimensions. This incoherent OCDMA approach is compatible with existing WDM optical networks and utilizes off the shelf components. We discuss the novel optical subsystem design for encoders and decoders that enable the realization of a highly scalable incoherent OCDMA system with rapid reconfigurability. A detailed analysis of the scalability of the two dimensional code is presented and select network deployment architectures for OCDMA are discussed (Authors)

  15. A Critical Appraisal of the Juvenile Justice System under Cameroon's 2005 Criminal Procedure Code: Emerging Challenges

    Directory of Open Access Journals (Sweden)

    S Tabe

    2012-03-01

    Full Text Available The objective of this article is to examine the changes introduced by the 2005 Cameroonian Criminal Procedure Code on matters of juvenile justice, considering that before this Code, juvenile justice in Cameroon was governed by extra-national laws. In undertaking this analysis, the article highlights the evolution of the administration of juvenile justice 50 years after independence of Cameroon. It also points out the various difficulties and shortcomings in the treatment of juvenile offenders in Cameroon since the enactment of the new Criminal Procedure Code. The article reveals that the 2005 Code is an amalgamation of all hitherto existing laws in the country that pertained to juvenile justice, and that despite the considerable amount of criticism it has received, the Code is clearly an improvement of the system of juvenile justice in Cameroon, since it represents a balance of the due process rights of young people, the protection of society and the special needs of young offenders. This is so because the drafters of the Code took a broad view of the old laws on juvenile justice. Also a wide range of groups were consulted, including criminal justice professionals, children’s service organisations, victims, parents, young offenders, educators, advocacy groups and social-policy analysts. However, to address the challenges that beset the juvenile justice system of Cameroon, the strategy of the government should be focussed on three areas: the prevention of youth crime, the provision of meaningful consequences for the actions of young people, and the rehabilitation and reintegration of young offenders. Cameroonian law should seek educative solutions rather than to impose prison sentences or other repressive measures on young offenders. Special courts to deal with young offenders should be established outside the regular penal system and should be provided with resources that are adequate for and appropriate to fostering their understanding of

  16. Overview of Particle and Heavy Ion Transport Code System PHITS

    Science.gov (United States)

    Sato, Tatsuhiko; Niita, Koji; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Furuta, Takuya; Noda, Shusaku; Ogawa, Tatsuhiko; Iwase, Hiroshi; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Chiba, Satoshi; Sihver, Lembit

    2014-06-01

    A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development's Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1,000 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.

  17. PERFORMANCE EVALUATION OF LOW DENSITY PARITY CHECK CODES FOR DIGITAL RADIO MONDIALE (DRM) SYSTEM

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In Digital Radio Mondiale (DRM) system, achieving good audio quality becomes a challenge due to its limited band-width of 9 or 10kHz and the very bad fading channels. Therefore, DRM needs highly efficient channel coding schemes. This paper, proposes the schemes which use the Low-Density Parity-Check (LDPC) coded Bit-Interleaved Coded Modulation (BICM) schemes for the implementation of DRM systems.Simulation results show that the proposed system is more efficient than the Rate Compatible Punctured Convolutional (RCPC) coded DRM system on various broadcast channels, and may be recommended as a coding technology for Digital Amplitude Modulation Broadcasting (DAMB) systems of China.

  18. Analysis of the optimality of the standard genetic code.

    Science.gov (United States)

    Kumar, Balaji; Saini, Supreet

    2016-07-19

    Many theories have been proposed attempting to explain the origin of the genetic code. While strong reasons remain to believe that the genetic code evolved as a frozen accident, at least for the first few amino acids, other theories remain viable. In this work, we test the optimality of the standard genetic code against approximately 17 million genetic codes, and locate 29 which outperform the standard genetic code at the following three criteria: (a) robustness to point mutation; (b) robustness to frameshift mutation; and (c) ability to encode additional information in the coding region. We use a genetic algorithm to generate and score codes from different parts of the associated landscape, which are, as a result, presumably more representative of the entire landscape. Our results show that while the genetic code is sub-optimal for robustness to frameshift mutation and the ability to encode additional information in the coding region, it is very strongly selected for robustness to point mutation. This coupled with the observation that the different performance indicator scores for a particular genetic code are negatively correlated makes the standard genetic code nearly optimal for the three criteria tested in this work. PMID:27327359

  19. Adjoint-based sensitivity analysis for reactor accident codes

    International Nuclear Information System (INIS)

    This paper summarizes a recently completed study that identified and investigated the difficulties and limitations of applying first-order adjoint sensitivity methods to reactor accident codes. The work extends earlier adjoint sensitivity formulations and applications to consider problem/model discontinuities in a general fashion, provide for response (R) formulations required by reactor safety applications, and provide a scheme for accurately handling extremely time-sensitive reactor accident responses. The scheme involves partitioning (dividing) the model into submodels (with spearate defining equations and initial conditions) at the location of discontinuity. Successful partitioning moves the problem dependence on the discontinuity location from the whole model system equations to the initial conditions of the second submodel

  20. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.