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Sample records for consequence code system

  1. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T.; Rollstin, J.A.; Chanin, D.I.

    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

  2. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Rollstin, J.A.; Chanin, D.I.; Jow, H.N.

    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 projections, 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

  3. 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.

  4. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Chanin, D.I.; Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian

    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

  5. 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.

  6. Accident consequence assessment code development

    International Nuclear Information System (INIS)

    Homma, T.; Togawa, O.

    1991-01-01

    This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)

  7. Study on the code system for the off-site consequences assessment of severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

  8. Study on the code system for the off-site consequences assessment of severe nuclear accident

    International Nuclear Information System (INIS)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents

  9. Quality assurance and verification of the MACCS [MELCOR Accident Consequence Code System] code, Version 1.5

    International Nuclear Information System (INIS)

    Dobbe, C.A.; Carlson, E.R.; Marshall, N.H.; Marwil, E.S.; Tolli, J.E.

    1990-02-01

    An independent quality assurance (QA) and verification of Version 1.5 of the MELCOR Accident Consequence Code System (MACCS) was performed. The QA and verification involved examination of the code and associated documentation for consistent and correct implementation of the models in an error-free FORTRAN computer code. The QA and verification was not intended to determine either the adequacy or appropriateness of the models that are used MACCS 1.5. The reviews uncovered errors which were fixed by the SNL MACCS code development staff prior to the release of MACCS 1.5. Some difficulties related to documentation improvement and code restructuring are also presented. The QA and verification process concluded that Version 1.5 of the MACCS code, within the scope and limitations process concluded that Version 1.5 of the MACCS code, within the scope and limitations of the models implemented in the code is essentially error free and ready for widespread use. 15 refs., 11 tabs

  10. A review of the Melcor Accident Consequence Code System (MACCS): Capabilities and applications

    International Nuclear Information System (INIS)

    Young, M.

    1995-01-01

    MACCS was developed at Sandia National Laboratories (SNL) under U.S. Nuclear Regulatory Commission (NRC) sponsorship to estimate the offsite consequences of potential severe accidents at nuclear power plants (NPPs). MACCS was publicly released in 1990. MACCS was developed to support the NRC's probabilistic safety assessment (PSA) efforts. PSA techniques can provide a measure of the risk of reactor operation. PSAs are generally divided into three levels. Level one efforts identify potential plant damage states that lead to core damage and the associated probabilities, level two models damage progression and containment strength for establishing fission-product release categories, and level three efforts evaluate potential off-site consequences of radiological releases and the probabilities associated with the consequences. MACCS was designed as a tool for level three PSA analysis. MACCS performs probabilistic health and economic consequence assessments of hypothetical accidental releases of radioactive material from NPPs. MACCS includes models for atmospheric dispersion and transport, wet and dry deposition, the probabilistic treatment of meteorology, environmental transfer, countermeasure strategies, dosimetry, health effects, and economic impacts. The computer systems MACCS is designed to run on are the 386/486 PC, VAX/VMS, E3M RISC S/6000, Sun SPARC, and Cray UNICOS. This paper provides an overview of MACCS, reviews some of the applications of MACCS, international collaborations which have involved MACCS, current developmental efforts, and future directions

  11. A parametric study of MELCOR Accident Consequence Code System 2 (MACCS2) Input Values for the Predicted Health Effect

    International Nuclear Information System (INIS)

    Kim, So Ra; Min, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The MELCOR Accident Consequence Code System 2, MACCS2, has been the most widely used through the world among the off-site consequence analysis codes. MACCS2 code is used to estimate the radionuclide concentrations, radiological doses, health effects, and economic consequences that could result from the hypothetical nuclear accidents. Most of the MACCS model parameter values are defined by the user and those input parameters can make a significant impact on the output. A limited parametric study was performed to identify the relative importance of the values of each input parameters in determining the predicted early and latent health effects in MACCS2. These results would not be applicable to every case of the nuclear accidents, because only the limited calculation was performed with Kori-specific data. The endpoints of the assessment were early- and latent cancer-risk in the exposed population, therefore it might produce the different results with the parametric studies for other endpoints, such as contamination level, absorbed dose, and economic cost. Accident consequence assessment is important for decision making to minimize the health effect from radiation exposure, accordingly the sufficient parametric studies are required for the various endpoints and input parameters in further research

  12. Evaluation of severe accident risks: Quantification of major input parameters: MAACS [MELCOR Accident Consequence Code System] input

    International Nuclear Information System (INIS)

    Sprung, J.L.; Jow, H-N; Rollstin, J.A.; Helton, J.C.

    1990-12-01

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs

  13. PROBCON-HDW: A probability and consequence system of codes for long-term analysis of Hanford defense wastes

    International Nuclear Information System (INIS)

    Piepho, M.G.; Nguyen, T.H.

    1988-12-01

    The PROBCON-HDW (PROBability and CONsequence analysis for Hanford defense waste) computer code system calculates the long-term cumulative releases of radionuclides from the Hanford defense wastes (HDW) to the accessible environment and compares the releases to environmental release limits as defined in 40 CFR 191. PROBCON-HDW takes into account the variability of input parameter values used in models to calculate HDW release and transport in the vadose zone to the accessible environment (taken here as groundwater). A human intrusion scenario, which consists of drilling boreholes into the waste beneath the waste sites and bringing waste to the surface, is also included in PROBCON-HDW. PROBCON-HDW also includes the capability to combine the cumulative releases according to various long-term (10,000 year) scenarios into a composite risk curve or complementary cumulative distribution function (CCDF). The system structure of the PROBCON-HDW codes, the mathematical models in PROBCON-HDW, the input files, the input formats, the command files, and the graphical output results of several HDW release scenarios are described in the report. 3 refs., 7 figs., 9 tabs

  14. [Quality management and strategic consequences of assessing documentation and coding under the German Diagnostic Related Groups system].

    Science.gov (United States)

    Schnabel, M; Mann, D; Efe, T; Schrappe, M; V Garrel, T; Gotzen, L; Schaeg, M

    2004-10-01

    The introduction of the German Diagnostic Related Groups (D-DRG) system requires redesigning administrative patient management strategies. Wrong coding leads to inaccurate grouping and endangers the reimbursement of treatment costs. This situation emphasizes the roles of documentation and coding as factors of economical success. The aims of this study were to assess the quantity and quality of initial documentation and coding (ICD-10 and OPS-301) and find operative strategies to improve efficiency and strategic means to ensure optimal documentation and coding quality. In a prospective study, documentation and coding quality were evaluated in a standardized way by weekly assessment. Clinical data from 1385 inpatients were processed for initial correctness and quality of documentation and coding. Principal diagnoses were found to be accurate in 82.7% of cases, inexact in 7.1%, and wrong in 10.1%. Effects on financial returns occurred in 16%. Based on these findings, an optimized, interdisciplinary, and multiprofessional workflow on medical documentation, coding, and data control was developed. Workflow incorporating regular assessment of documentation and coding quality is required by the DRG system to ensure efficient accounting of hospital services. Interdisciplinary and multiprofessional cooperation is recognized to be an important factor in establishing an efficient workflow in medical documentation and coding.

  15. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  16. Review of the chronic exposure pathways models in MACCS [MELCOR Accident Consequence Code System] and several other well-known probabilistic risk assessment models

    International Nuclear Information System (INIS)

    Tveten, U.

    1990-06-01

    The purpose of this report is to document the results of the work performed by the author in connection with the following task, performed for US Nuclear Regulatory Commission, (USNRC) Office of Nuclear Regulatory Research, Division of Systems Research: MACCS Chronic Exposure Pathway Models: Review the chronic exposure pathway models implemented in the MELCOR Accident Consequence Code System (MACCS) and compare those models to the chronic exposure pathway models implemented in similar codes developed in countries that are members of the OECD. The chronic exposures concerned are via: the terrestrial food pathways, the water pathways, the long-term groundshine pathway, and the inhalation of resuspended radionuclides pathway. The USNRC has indicated during discussions of the task that the major effort should be spent on the terrestrial food pathways. There is one chapter for each of the categories of chronic exposure pathways listed above

  17. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    Togawa, O.; Homma, T.

    1992-01-01

    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)

  18. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    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

  19. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  20. Analyses of PWR boron dilution consequences with the Arrotta code

    International Nuclear Information System (INIS)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R.

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced

  1. Revised SRAC code system

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.

    1986-09-01

    Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)

  2. 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

  3. EDPUFF- a Gaussian dispersion code for consequence analysis

    International Nuclear Information System (INIS)

    Oza, R.B.; Bapat, V.N.; Nair, R.N.; Hukkoo, R.K.; Krishnamoorthy, T.M.

    1995-01-01

    EDPUFF- Equi Distance Puff is a Gaussian dispersion code in FORTRAN language to model atmospheric dispersion of instantaneous or continuous point source releases. It is designed to incorporate the effect of changing meteorological conditions and source release rates on the spatial distribution profiles and its consequences. Effects of variation of parameters like puff spacing, puff packing, averaging schemes are discussed and the choice of the best values for minimum errors and minimum computer CPU time are identified. The code calculates the doses to individual receptors as well as average doses for population zones from internal and external routes over the area of interest. Internal dose computations are made for inhalation and ingestion pathways while the doses from external route consists of cloud doses and doses from surface deposited activity. It computes inhalation and ingestion dose (milk route only) for critical group (1 yr old child). In case of population zones it finds out maximum possible doses in a given area along with the average doses discussed above. Report gives the doses from various pathways for unit release of fixed duration. (author). 7 refs., figs., 7 appendixes

  4. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  5. Elements of algebraic coding systems

    CERN Document Server

    Cardoso da Rocha, Jr, Valdemar

    2014-01-01

    Elements of Algebraic Coding Systems is an introductory text to algebraic coding theory. In the first chapter, you'll gain inside knowledge of coding fundamentals, which is essential for a deeper understanding of state-of-the-art coding systems. This book is a quick reference for those who are unfamiliar with this topic, as well as for use with specific applications such as cryptography and communication. Linear error-correcting block codes through elementary principles span eleven chapters of the text. Cyclic codes, some finite field algebra, Goppa codes, algebraic decoding algorithms, and applications in public-key cryptography and secret-key cryptography are discussed, including problems and solutions at the end of each chapter. Three appendices cover the Gilbert bound 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 finite field Fourier transform and the Euclidean algorithm f...

  6. SCALE Code System

    Energy Technology Data Exchange (ETDEWEB)

    Jessee, Matthew Anderson [ORNL

    2016-04-01

    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.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE

  7. System Based Code: Principal Concept

    International Nuclear Information System (INIS)

    Yasuhide Asada; Masanori Tashimo; Masahiro Ueta

    2002-01-01

    This paper introduces a concept of the 'System Based Code' which has initially been proposed by the authors intending to give nuclear industry a leap of progress in the system reliability, performance improvement, and cost reduction. The concept of the System Based Code intends to give a theoretical procedure to optimize the reliability of the system by administrating every related engineering requirement throughout the life of the system from design to decommissioning. (authors)

  8. Methods and codes for assessing the off-site Consequences of nuclear accidents. Volume 2

    International Nuclear Information System (INIS)

    Kelly, G.N.; Luykx, F.

    1991-01-01

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled methods for assessing the radiological impact of accidents (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  9. ESCADRE and ICARE code systems

    International Nuclear Information System (INIS)

    Reocreux, M.; Gauvain, J.

    1992-01-01

    The French sever accident code development program is following two parallel approaches: the first one is dealing with ''integral codes'' which are designed for giving immediate engineer answers, the second one is following a more mechanistic way in order to have the capability of detailed analysis of experiments, in order to get a better understanding of the scaling problem and reach a better confidence in plant calculations. In the first approach a complete system has been developed and is being used for practical cases: this is the ESCADRE system. In the second approach, a set of codes dealing first with primary circuit is being developed: a mechanistic core degradation code, ICARE, has been issued and is being coupled with the advanced thermalhydraulic code CATHARE. Fission product codes have been also coupled to CATHARE. The ''integral'' ESCADRE system and the mechanistic ICARE and associated codes are described. Their main characteristics are reviewed and the status of their development and assessment given. Future studies are finally discussed. 36 refs, 4 figs, 1 tab

  10. SPECTRAL AMPLITUDE CODING OCDMA SYSTEMS USING ENHANCED DOUBLE WEIGHT CODE

    Directory of Open Access Journals (Sweden)

    F.N. HASOON

    2006-12-01

    Full Text Available A new code structure for spectral amplitude coding optical code division multiple access systems based on double weight (DW code families is proposed. The DW has a fixed weight of two. Enhanced double-weight (EDW code is another variation of a DW code family that can has a variable weight greater than one. The EDW code possesses ideal cross-correlation properties and exists for every natural number n. A much better performance can be provided by using the EDW code compared to the existing code such as Hadamard and Modified Frequency-Hopping (MFH codes. It has been observed that theoretical analysis and simulation for EDW is much better performance compared to Hadamard and Modified Frequency-Hopping (MFH codes.

  11. 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.

  12. ETR/ITER systems code

    International Nuclear Information System (INIS)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.

    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

  13. Evaluation system for the analysis of the radiological consequences (RASCAL)

    International Nuclear Information System (INIS)

    Miguel Martinez, M. J. de

    2011-01-01

    The code currently employed by the Spanish Nuclear power is the IRDAM (Interactive Rapid Dose Assessment Model). This code will now be replaced by the RASCAL (Radiological Assessment System for Consequence Analysis) version 4.0. This is a significant improvement in dose calculation for escapes to the atmosphere for radioactive emergencies. The main objective is to highlight the most significant changes introduced in the RASCAL when performing these dose estimates.

  14. An uncertainty analysis using the NRPB accident consequence code Marc

    International Nuclear Information System (INIS)

    Jones, J.A.; Crick, M.J.; Simmonds, J.R.

    1991-01-01

    This paper describes an uncertainty analysis of MARC calculations of the consequences of accidental releases of radioactive materials to atmosphere. A total of 98 parameters describing the transfer of material through the environment to man, the doses received, and the health effects resulting from these doses, was considered. The uncertainties in the numbers of early and late health effects, numbers of people affected by countermeasures, the amounts of food restricted and the economic costs of the accident were estimated. This paper concentrates on the results for early death and fatal cancer for a large hypothetical release from a PWR

  15. Interrelations of codes in human semiotic systems.

    OpenAIRE

    Somov, Georgij

    2016-01-01

    Codes can be viewed as mechanisms that enable relations of signs and their components, i.e., semiosis is actualized. The combinations of these relations produce new relations as new codes are building over other codes. Structures appear in the mechanisms of codes. Hence, codes can be described as transformations of structures from some material systems into others. Structures belong to different carriers, but exist in codes in their "pure" form. Building of codes over other codes fosters t...

  16. System Design Description for the TMAD Code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the System Design Description (SDD) for the TMAD Code System, which includes the TMAD code and the LIBMAKR code. The SDD provides a detailed description of the theory behind the code, and the implementation of that theory. It is essential for anyone who is attempting to review or modify the code or who otherwise needs to understand the internal workings of the code. In addition, this document includes, in Appendix A, the System Requirements Specification for the TMAD System

  17. General Logic-Systems and Consequence Operators

    OpenAIRE

    Herrmann, Robert A.

    2005-01-01

    In this paper, general logic-systems are investigated. It is shown that there are infinitely many finite consequence operators defined on a fixed language L that cannot be generated from a finite logic-system. It is shown that a set map is a finite consequence operator iff it is defined by a general logic-system.

  18. 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

  19. 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.

  20. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  1. Expansion of the CHR bone code system

    International Nuclear Information System (INIS)

    Farnham, J.E.; Schlenker, R.A.

    1976-01-01

    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

  2. Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

    International Nuclear Information System (INIS)

    Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

    2002-01-01

    A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

  3. Coding-Spreading Tradeoff in CDMA Systems

    National Research Council Canada - National Science Library

    Bolas, Eduardo

    2002-01-01

    .... Comparing different combinations of coding and spreading with a traditional DS-CDMA, as defined in the IS-95 standard, allows the criteria to be defined for the best coding-spreading tradeoff in CDMA systems...

  4. Parameterization of the driving time in the evacuation or fast relocation model of an accident consequence code

    International Nuclear Information System (INIS)

    Pfeffer, W.; Hofer, E.; Nowak, E.; Schnadt, H.

    1988-01-01

    The model of protective measures in the accident consequence code system UFOMOD of the German Risk Study, Phase B, requires the driving times of the population to be evacuated for the evaluation of the dose received during the evacuation. The parameter values are derived from evacuation simulations carried out with the code EVAS for 36 sectors from various sites. The simulations indicated that the driving time strongly depends on the population density, whereas other influences are less important. It was decided to use different driving times in the consequence code for each of four population density classes as well as for each of three or four fractions of the population in a sector. The variability between sectors of a class was estimated from the 36 sectors, in order to derive subjective probability distributions that are to model the uncertainty in the parameter value to be used for any of the fractions in a particular sector for which an EVAS simulation has not yet been performed. To this end also the impact of the uncertainties in the parameters and modelling assumptions of EVAS on the simulated times was quantified using expert judgement. The distributions permit the derivation of a set of driving times to be used as so-called ''best estimate'' or reference values in the accident consequence code. Additionally they are directly applicable in an uncertainty and sensitivity analysis

  5. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  6. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  7. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  8. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

  9. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    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

  10. Experience with COSYMA in an international intercomparison of probabilistic accident consequence assessment codes

    International Nuclear Information System (INIS)

    Hasemann, I.; Jones, J.A.; Steen, J. van der; Wonderen, E. van

    1996-01-01

    The Commission of the European Communities and the Nuclear Energy Agency of the OECD have organized an international exercise to compare the predictions of accident consequence assessment codes, and to identify those features of the models which lead to differences in the predicted results. Alongside this, a further exercise was undertaken in which the COSYMA code was used independently by several different organizations. Some of the findings of the COSYMA users' exercise are described that have general applications to accident consequence assessments. A number of areas are identified in which further work on accident consequence models may be justified. These areas, which are also of interest for codes other than COSYMA, are (a) the calculation and averaging of doses and risks to people sheltered in different types of buildings, particularly with respect to the evaluation of early health effects; (b) the modeling of long-duration releases and their description as a series of shorter releases; (c) meteorological sampling for results at a certain location, specifically for use with trajectory models of atmospheric dispersion; and (d) aspects of calculating probabilities of consequences at a point

  11. Proceedings of the Seminar on Methods and Codes for Assessing the off-site consequences of nuclear accidents. Volume 1

    International Nuclear Information System (INIS)

    Kelly, G.N.; Luykx, F.

    1991-01-01

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled 'methods for assessing the radiological impact of accidents' (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  12. High dynamic range coding imaging system

    Science.gov (United States)

    Wu, Renfan; Huang, Yifan; Hou, Guangqi

    2014-10-01

    We present a high dynamic range (HDR) imaging system design scheme based on coded aperture technique. This scheme can help us obtain HDR images which have extended depth of field. We adopt Sparse coding algorithm to design coded patterns. Then we utilize the sensor unit to acquire coded images under different exposure settings. With the guide of the multiple exposure parameters, a series of low dynamic range (LDR) coded images are reconstructed. We use some existing algorithms to fuse and display a HDR image by those LDR images. We build an optical simulation model and get some simulation images to verify the novel system.

  13. Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Paik, I.K.; Chung, D.Y.

    1996-01-01

    Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, including source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper

  14. Confidence level in the calculations of HCDA consequences using large codes

    International Nuclear Information System (INIS)

    Nguyen, D.H.; Wilburn, N.P.

    1979-01-01

    The probabilistic approach to nuclear reactor safety is playing an increasingly significant role. For the liquid-metal fast breeder reactor (LMFBR) in particular, the ultimate application of this approach could be to determine the probability of achieving the goal of a specific line-of-assurance (LOA). Meanwhile a more pressing problem is one of quantifying the uncertainty in a calculated consequence for hypothetical core disruptive accident (HCDA) using large codes. Such uncertainty arises from imperfect modeling of phenomenology and/or from inaccuracy in input data. A method is presented to determine the confidence level in consequences calculated by a large computer code due to the known uncertainties in input invariables. A particular application was made to the initial time of pin failure in a transient overpower HCDA calculated by the code MELT-IIIA in order to demonstrate the method. A probability distribution function (pdf) for the time of failure was first constructed, then the confidence level for predicting this failure parameter within a desired range was determined

  15. Advanced thermionic reactor systems design code

    International Nuclear Information System (INIS)

    Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

    1991-01-01

    An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

  16. ETF system code: composition and applications

    International Nuclear Information System (INIS)

    Reid, R.L.; Wu, K.F.

    1980-01-01

    A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies, such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system

  17. A bar coding system for environmental projects

    International Nuclear Information System (INIS)

    Barber, R.B.; Hunt, B.J.; Burgess, G.M.

    1988-01-01

    This paper presents BeCode systems, a bar coding system which provides both nuclear and commercial clients with a data capture and custody management program that is accurate, timely, and beneficial to all levels of project operations. Using bar code identifiers is an essentially paperless and error-free method which provides more efficient delivery of data through its menu card-driven structure, which speeds collection of essential data for uploading to a compatible device. The effects of this sequence include real-time information for operator analysis, management review, audits, planning, scheduling, and cost control

  18. Arabic Natural Language Processing System Code Library

    Science.gov (United States)

    2014-06-01

    Adelphi, MD 20783-1197 This technical note provides a brief description of a Java library for Arabic natural language processing ( NLP ) containing code...for training and applying the Arabic NLP system described in the paper "A Cross-Task Flexible Transition Model for Arabic Tokenization, Affix...and also English) natural language processing ( NLP ), containing code for training and applying the Arabic NLP system described in Stephen Tratz’s

  19. User effects on the transient system code calculations. Final report

    International Nuclear Information System (INIS)

    Aksan, S.N.; D'Auria, F.

    1995-01-01

    Large thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermal-hydraulic transient system codes is presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects are discussed in detail and general recommendations and conclusions are presented to control and limit them

  20. Evaluation of dose calculation models for inhabited areas applicable in nuclear accident consequence assessment codes

    International Nuclear Information System (INIS)

    Katalin Eged; Zoltan Kis; Natalia Semioschkina; Gabriele Voigt

    2004-01-01

    One of the objectives of the EC project EVANET-TERRA is to provide suitable inputs to the RODOS system. This study gives an overview on urban dose calculation models with special emphasis on the RECLAIM-EDEM2M and TEMAS-urban codes. The TEMAS-urban code is more complex compared to the RECLAIM-EDEM2M code although both models use similar and some times even same model parameters. The database and the way of its data collection as used in RECLAIM-EDEM2M is recommended as a preferred option because it contains many data from local and regional measurements. However in a decision situation the outputs of the TEMASurban model may better help stake holders by providing a ranking of the surfaces to be decontaminated. (author)

  1. Fast decoding algorithms for coded aperture systems

    International Nuclear Information System (INIS)

    Byard, Kevin

    2014-01-01

    Fast decoding algorithms are described for a number of established coded aperture systems. The fast decoding algorithms for all these systems offer significant reductions in the number of calculations required when reconstructing images formed by a coded aperture system and hence require less computation time to produce the images. The algorithms may therefore be of use in applications that require fast image reconstruction, such as near real-time nuclear medicine and location of hazardous radioactive spillage. Experimental tests confirm the efficacy of the fast decoding techniques

  2. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    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)

  3. Implementing a modular system of computer codes

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-07-01

    A modular computation system has been developed for nuclear reactor core analysis. The codes can be applied repeatedly in blocks without extensive user input data, as needed for reactor history calculations. The primary control options over the calculational paths and task assignments within the codes are blocked separately from other instructions, admitting ready access by user input instruction or directions from automated procedures and promoting flexible and diverse applications at minimum application cost. Data interfacing is done under formal specifications with data files manipulated by an informed manager. This report emphasizes the system aspects and the development of useful capability, hopefully informative and useful to anyone developing a modular code system of much sophistication. Overall, this report in a general way summarizes the many factors and difficulties that are faced in making reactor core calculations, based on the experience of the authors. It provides the background on which work on HTGR reactor physics is being carried out

  4. Plotting system for the MINCS code

    International Nuclear Information System (INIS)

    Watanabe, Tadashi

    1990-08-01

    The plotting system for the MINCS code is described. The transient two-phase flow analysis code MINCS has been developed to provide a computational tool for analysing various two-phase flow phenomena in one-dimensional ducts. Two plotting systems, namely the SPLPLOT system and the SDPLOT system, can be used as the plotting functions. The SPLPLOT system is used for plotting time transients of variables, while the SDPLOT system is for spatial distributions. The SPLPLOT system is based on the SPLPACK system, which is used as a general tool for plotting results of transient analysis codes or experiments. The SDPLOT is based on the GPLP program, which is also regarded as one of the general plotting programs. In the SPLPLOT and the SDPLOT systems, the standardized data format called the SPL format is used in reading data to be plotted. The output data format of MINCS is translated into the SPL format by using the conversion system called the MINTOSPL system. In this report, how to use the plotting functions is described. (author)

  5. Development of the versatile reactor analysis code system, MARBLE2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

    2015-07-01

    The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

  6. Improved decoding for a concatenated coding system

    DEFF Research Database (Denmark)

    Paaske, Erik

    1990-01-01

    The concatenated coding system recommended by CCSDS (Consultative Committee for Space Data Systems) uses an outer (255,233) Reed-Solomon (RS) code based on 8-b symbols, followed by the block interleaver and an inner rate 1/2 convolutional code with memory 6. Viterbi decoding is assumed. Two new...... decoding procedures based on repeated decoding trials and exchange of information between the two decoders and the deinterleaver are proposed. In the first one, where the improvement is 0.3-0.4 dB, only the RS decoder performs repeated trials. In the second one, where the improvement is 0.5-0.6 dB, both...... decoders perform repeated decoding trials and decoding information is exchanged between them...

  7. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  8. Symbol synchronization in convolutionally coded systems

    Science.gov (United States)

    Baumert, L. D.; Mceliece, R. J.; Van Tilborg, H. C. A.

    1979-01-01

    Alternate symbol inversion is sometimes applied to the output of convolutional encoders to guarantee sufficient richness of symbol transition for the receiver symbol synchronizer. A bound is given for the length of the transition-free symbol stream in such systems, and those convolutional codes are characterized in which arbitrarily long transition free runs occur.

  9. Coding and decoding for code division multiple user communication systems

    Science.gov (United States)

    Healy, T. J.

    1985-01-01

    A new algorithm is introduced which decodes code division multiple user communication signals. The algorithm makes use of the distinctive form or pattern of each signal to separate it from the composite signal created by the multiple users. Although the algorithm is presented in terms of frequency-hopped signals, the actual transmitter modulator can use any of the existing digital modulation techniques. The algorithm is applicable to error-free codes or to codes where controlled interference is permitted. It can be used when block synchronization is assumed, and in some cases when it is not. The paper also discusses briefly some of the codes which can be used in connection with the algorithm, and relates the algorithm to past studies which use other approaches to the same problem.

  10. Bar-code automated waste tracking system

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. Analog system for computing sparse codes

    Science.gov (United States)

    Rozell, Christopher John; Johnson, Don Herrick; Baraniuk, Richard Gordon; Olshausen, Bruno A.; Ortman, Robert Lowell

    2010-08-24

    A parallel dynamical system for computing sparse representations of data, i.e., where the data can be fully represented in terms of a small number of non-zero code elements, and for reconstructing compressively sensed images. The system is based on the principles of thresholding and local competition that solves a family of sparse approximation problems corresponding to various sparsity metrics. The system utilizes Locally Competitive Algorithms (LCAs), nodes in a population continually compete with neighboring units using (usually one-way) lateral inhibition to calculate coefficients representing an input in an over complete dictionary.

  13. A Consistent System for Coding Laboratory Samples

    Science.gov (United States)

    Sih, John C.

    1996-07-01

    A formal laboratory coding system is presented to keep track of laboratory samples. Preliminary useful information regarding the sample (origin and history) is gained without consulting a research notebook. Since this system uses and retains the same research notebook page number for each new experiment (reaction), finding and distinguishing products (samples) of the same or different reactions becomes an easy task. Using this system multiple products generated from a single reaction can be identified and classified in a uniform fashion. Samples can be stored and filed according to stage and degree of purification, e.g. crude reaction mixtures, recrystallized samples, chromatographed or distilled products.

  14. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

    1997-06-01

    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)

  15. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

    1997-06-01

    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 {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)

  16. A mean field theory of coded CDMA systems

    International Nuclear Information System (INIS)

    Yano, Toru; Tanaka, Toshiyuki; Saad, David

    2008-01-01

    We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems

  17. A mean field theory of coded CDMA systems

    Energy Technology Data Exchange (ETDEWEB)

    Yano, Toru [Graduate School of Science and Technology, Keio University, Hiyoshi, Kohoku-ku, Yokohama-shi, Kanagawa 223-8522 (Japan); Tanaka, Toshiyuki [Graduate School of Informatics, Kyoto University, Yoshida Hon-machi, Sakyo-ku, Kyoto-shi, Kyoto 606-8501 (Japan); Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)], E-mail: yano@thx.appi.keio.ac.jp

    2008-08-15

    We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems.

  18. SRAC95; general purpose neutronics code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Tsuchihashi, Keichiro; 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)

  19. 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).

  20. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

  1. Safety assessment of high consequence robotics system

    International Nuclear Information System (INIS)

    Robinson, D.G.; Atcitty, C.B.

    1996-01-01

    This paper outlines the use of a failure modes and effects analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, the weigh and leak check system, is to replace a manual process for weight and leakage of nuclear materials at the DOE Pantex facility. Failure modes and effects analyses were completed for the robotics process to ensure that safety goals for the systems have been met. Due to the flexible nature of the robot configuration, traditional failure modes and effects analysis (FMEA) were not applicable. In addition, the primary focus of safety assessments of robotics systems has been the protection of personnel in the immediate area. In this application, the safety analysis must account for the sensitivities of the payload as well as traditional issues. A unique variation on the classical FMEA was developed that permits an organized and quite effective tool to be used to assure that safety was adequately considered during the development of the robotic system. The fundamental aspects of the approach are outlined in the paper

  2. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

  3. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

  4. Apply Functional Modelling to Consequence Analysis in Supervision Systems

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Lind, Morten; Gola, Giulio

    2013-01-01

    This paper will first present the purpose and goals of applying functional modelling approach to consequence analysis by adopting Multilevel Flow Modelling (MFM). MFM Models describe a complex system in multiple abstraction levels in both means-end dimension and whole-part dimension. It contains...... consequence analysis to practical or online applications in supervision systems. It will also suggest a multiagent solution as the integration architecture for developing tools to facilitate the utilization results of functional consequence analysis. Finally a prototype of the multiagent reasoning system...... causal relations between functions and goals. A rule base system can be developed to trace the causal relations and perform consequence propagations. This paper will illustrate how to use MFM for consequence reasoning by using rule base technology and describe the challenges for integrating functional...

  5. Participation in the international comparison of probabilistic consequence assessment codes organized by OECD/NEA and CEC. Final Report

    International Nuclear Information System (INIS)

    Rossi, J.

    1994-02-01

    Probabilistic Consequence Assessment (PCA) methods are exploited not only in risk evaluation but also to study alternative design features, reactor siting recommendations and to obtain acceptable dose criteria by the radiation safety authorities. The models are programmed into computer codes for these kind of assessment. To investigate the quality and competence of different models, OECD/NEA and CEC organized the international code comparison exercise, which was participated by the organizations from 15 countries. There were seven codes participating in the exercise. The objectives of the code comparison exercise were to compare the results by the codes, to contribute to PCA code quality assurance, to harmonize the codes, to provide a forum for discussion on various approaches and to produce the report on the exercise. The project started in 1991 and the results of the calculations were completed in autumn 1992. The international report consists of two parts: the Overview Report for decision makers and the supporting detailed Technical Report. The results of the project are reviewed as an user of the ARANO-programme of VTT and trends of it's further development are indicated in this report. (orig.) (11 refs., 13 figs., 4 tabs.)

  6. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant 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 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 core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user 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 functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; 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. For this

  7. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

    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 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 criticality 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 characteristics 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. For

  8. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1979-03-01

    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)

  9. Concatenated coding system with iterated sequential inner decoding

    DEFF Research Database (Denmark)

    Jensen, Ole Riis; Paaske, Erik

    1995-01-01

    We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...

  10. Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

    International Nuclear Information System (INIS)

    Neymotin, L.

    1994-04-01

    Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions

  11. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1988-01-01

    This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared

  12. RASCAL 4.l: evaluation system for the analysis of the radiological consequences

    International Nuclear Information System (INIS)

    Miguel, I. de; Gomez-Arguello, B.

    2012-01-01

    Recently the Spanish Nuclear Safety Council (CSN) has promoted the replacement of the code used in the Spanish Nuclear Power Plants to estimate potential and actual doses due to atmospheric releases. IRDAM Code (interactive Rapid Dose Assessment Model, 1983) by RASCAL Code version 4.1 (Radiological Assessment System for consequence Analysis). This code was developed by the NRC (Nuclear Regulatory Commission) to evaluate releases from nuclear power plants, spent fuel storage pools and casks, fuel cycle facilities and radioactive material handling facilities. RASCAL was officially requested to the Spanish nuclear power plants in February 1st, 2012, and it can be used for external dose assessment and to evaluate the recommended protective action for the public in a specific situation (sheltering, prophylaxis, evacuation, etc.). (Author)

  13. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  14. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  15. Optical code division multiple access secure communications systems with rapid reconfigurable polarization shift key user code

    Science.gov (United States)

    Gao, Kaiqiang; Wu, Chongqing; Sheng, Xinzhi; Shang, Chao; Liu, Lanlan; Wang, Jian

    2015-09-01

    An optical code division multiple access (OCDMA) secure communications system scheme with rapid reconfigurable polarization shift key (Pol-SK) bipolar user code is proposed and demonstrated. Compared to fix code OCDMA, by constantly changing the user code, the performance of anti-eavesdropping is greatly improved. The Pol-SK OCDMA experiment with a 10 Gchip/s user code and a 1.25 Gb/s user data of payload has been realized, which means this scheme has better tolerance and could be easily realized.

  16. Status of reactor core design code system in COSINE code package

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Liu, Z.

    2014-01-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  17. Status of reactor core design code system in COSINE code package

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

    2014-07-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  18. 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.

  19. Joint design of QC-LDPC codes for coded cooperation system with joint iterative decoding

    Science.gov (United States)

    Zhang, Shunwai; Yang, Fengfan; Tang, Lei; Ejaz, Saqib; Luo, Lin; Maharaj, B. T.

    2016-03-01

    In this paper, we investigate joint design of quasi-cyclic low-density-parity-check (QC-LDPC) codes for coded cooperation system with joint iterative decoding in the destination. First, QC-LDPC codes based on the base matrix and exponent matrix are introduced, and then we describe two types of girth-4 cycles in QC-LDPC codes employed by the source and relay. In the equivalent parity-check matrix corresponding to the jointly designed QC-LDPC codes employed by the source and relay, all girth-4 cycles including both type I and type II are cancelled. Theoretical analysis and numerical simulations show that the jointly designed QC-LDPC coded cooperation well combines cooperation gain and channel coding gain, and outperforms the coded non-cooperation under the same conditions. Furthermore, the bit error rate performance of the coded cooperation employing jointly designed QC-LDPC codes is better than those of random LDPC codes and separately designed QC-LDPC codes over AWGN channels.

  20. A radiological accident consequence assessment system for Hong Kong

    International Nuclear Information System (INIS)

    Wong, M.C.; Lam, H.K.

    1993-01-01

    An account is given of the Hong Kong Radiological Accident Consequence Assessment System which would be used to assess the potential consequences of an emergency situation involving atmospheric release of radioactive material. The system has the capability to acquire real-time meteorological information from the Observatory's network of automatic stations, synoptic stations in the nearby region as well as forecast data from numerical prediction models. The system makes use of these data to simulate the transport and dispersion of the released radioactive material. The effectiveness of protective action on the local population is also modeled. The system serves as a powerful aid in the protective action recommendation processes

  1. The consequences of multiplexing and limited view angle in coded-aperture imaging

    International Nuclear Information System (INIS)

    Smith, W.E.; Barrett, H.H.; Paxman, R.G.

    1984-01-01

    Coded-aperture imaging (CAI) is a method for reconstructing distributions of radionuclide tracers that offers advantages over ECT and PET; namely, many views can be taken simultaneously without detector motion, and large numbers of photons are utilized since collimators are not required. However, because of this type of data acquisition, the coded image suffers from multiplexing; i.e., more than one object point may be mapped to each detector in the coded image. To investigate the dependence of the reconstruction on multiplexing, the authors reconstruct a simulated two-dimensional circular object from multiplexed one-dimensional coded-image data, then perform the reconstruction from un-multiplexed data. Each of these reconstructions are produced both from noise-free and noisy simulated data. To investigate the dependence on view angle, the authors reconstruct two simulated three-dimensional objects; a spherical phantom, and a series of point-like objects arranged nearly in a plane. Each of these reconstructions are from multiplexed two-dimensional coded-image data, first using two orthogonal views, and then a single viewing direction. The two-dimensional reconstructions demonstrate that, in the noise-free case, the multiplexing of the data does not seriously affect the reconstruction equality and that in the noisy-data case, the multiplexing helps, due to the fact that more photons are collected. Also, for point-like objects confined to a near-planar region of space, the authors show that restricted views can give satisfactory results, but that, for a large, three-dimensional object, a more complete viewing geometry is required

  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. SCALE Code System 6.2.2

    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)

    2017-05-01

    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 physics, radiation shielding, 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 3 deterministic and 3 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. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.

  4. SCALE Code System 6.2.1

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Jessee, Matthew Anderson

    2016-01-01

    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.

  5. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  6. 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.

  7. Next generation Zero-Code control system UI

    CERN Multimedia

    CERN. Geneva

    2017-01-01

    Developing ergonomic user interfaces for control systems is challenging, especially during machine upgrade and commissioning where several small changes may suddenly be required. Zero-code systems, such as *Inspector*, provide agile features for creating and maintaining control system interfaces. More so, these next generation Zero-code systems bring simplicity and uniformity and brake the boundaries between Users and Developers. In this talk we present *Inspector*, a CERN made Zero-code application development system, and we introduce the major differences and advantages of using Zero-code control systems to develop operational UI.

  8. Variable-length code construction for incoherent optical CDMA systems

    Science.gov (United States)

    Lin, Jen-Yung; Jhou, Jhih-Syue; Wen, Jyh-Horng

    2007-04-01

    The purpose of this study is to investigate the multirate transmission in fiber-optic code-division multiple-access (CDMA) networks. In this article, we propose a variable-length code construction for any existing optical orthogonal code to implement a multirate optical CDMA system (called as the multirate code system). For comparison, a multirate system where the lower-rate user sends each symbol twice is implemented and is called as the repeat code system. The repetition as an error-detection code in an ARQ scheme in the repeat code system is also investigated. Moreover, a parallel approach for the optical CDMA systems, which is proposed by Marić et al., is also compared with other systems proposed in this study. Theoretical analysis shows that the bit error probability of the proposed multirate code system is smaller than other systems, especially when the number of lower-rate users is large. Moreover, if there is at least one lower-rate user in the system, the multirate code system accommodates more users than other systems when the error probability of system is set below 10 -9.

  9. A Robust Cross Coding Scheme for OFDM Systems

    NARCIS (Netherlands)

    Shao, X.; Slump, Cornelis H.

    2010-01-01

    In wireless OFDM-based systems, coding jointly over all the sub-carriers simultaneously performs better than coding separately per sub-carrier. However, the joint coding is not always optimal because its achievable channel capacity (i.e. the maximum data rate) is inversely proportional to the

  10. Variable code gamma ray imaging system

    International Nuclear Information System (INIS)

    Macovski, A.; Rosenfeld, D.

    1979-01-01

    A gamma-ray source distribution in the body is imaged onto a detector using an array of apertures. The transmission of each aperture is modulated using a code such that the individual views of the source through each aperture can be decoded and separated. The codes are chosen to maximize the signal to noise ratio for each source distribution. These codes determine the photon collection efficiency of the aperture array. Planar arrays are used for volumetric reconstructions and circular arrays for cross-sectional reconstructions. 14 claims

  11. Channel coding in the space station data system network

    Science.gov (United States)

    Healy, T.

    1982-01-01

    A detailed discussion of the use of channel coding for error correction, privacy/secrecy, channel separation, and synchronization is presented. Channel coding, in one form or another, is an established and common element in data systems. No analysis and design of a major new system would fail to consider ways in which channel coding could make the system more effective. The presence of channel coding on TDRS, Shuttle, the Advanced Communication Technology Satellite Program system, the JSC-proposed Space Operations Center, and the proposed 30/20 GHz Satellite Communication System strongly support the requirement for the utilization of coding for the communications channel. The designers of the space station data system have to consider the use of channel coding.

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

    International Nuclear Information System (INIS)

    Forster, R.A.; Parsons, K.

    1997-01-01

    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

  13. A Cause-Consequence Chart of a Redundant Protection System

    DEFF Research Database (Denmark)

    Nielsen, Dan Sandvik; Platz, O.; Runge, B.

    1975-01-01

    A cause-consequence chart is applied in analysing failures of a redundant protection system (a core spray system in a nuclear power plant). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of the protection system. The test policy of compone...... of components is taken into account. The possibility of using parameter variation as a basis for the choice of test policy is indicated.......A cause-consequence chart is applied in analysing failures of a redundant protection system (a core spray system in a nuclear power plant). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of the protection system. The test policy...

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

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2002-11-01

    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)

  15. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2015-01-01

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed 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. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. 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 bundle can be reduced effectively by the safety measures

  16. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany)

    2015-04-15

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed 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. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. 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 bundle can be reduced effectively by the safety measures

  17. A Neural Code That Is Isometric to Vocal Output and Correlates with Its Sensory Consequences.

    Directory of Open Access Journals (Sweden)

    Alexei L Vyssotski

    2016-10-01

    Full Text Available What cortical inputs are provided to motor control areas while they drive complex learned behaviors? We study this question in the nucleus interface of the nidopallium (NIf, which is required for normal birdsong production and provides the main source of auditory input to HVC, the driver of adult song. In juvenile and adult zebra finches, we find that spikes in NIf projection neurons precede vocalizations by several tens of milliseconds and are insensitive to distortions of auditory feedback. We identify a local isometry between NIf output and vocalizations: quasi-identical notes produced in different syllables are preceded by highly similar NIf spike patterns. NIf multiunit firing during song precedes responses in auditory cortical neurons by about 50 ms, revealing delayed congruence between NIf spiking and a neural representation of auditory feedback. Our findings suggest that NIf codes for imminent acoustic events within vocal performance.

  18. The application of LDPC code in MIMO-OFDM system

    Science.gov (United States)

    Liu, Ruian; Zeng, Beibei; Chen, Tingting; Liu, Nan; Yin, Ninghao

    2018-03-01

    The combination of MIMO and OFDM technology has become one of the key technologies of the fourth generation mobile communication., which can overcome the frequency selective fading of wireless channel, increase the system capacity and improve the frequency utilization. Error correcting coding introduced into the system can further improve its performance. LDPC (low density parity check) code is a kind of error correcting code which can improve system reliability and anti-interference ability, and the decoding is simple and easy to operate. This paper mainly discusses the application of LDPC code in MIMO-OFDM system.

  19. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro

    2007-02-01

    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, S N 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)

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

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Yoon, Churl

    2010-02-01

    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

  1. Selected systems engineering process deficiencies and their consequences

    Science.gov (United States)

    Thomas, L. Dale

    2007-06-01

    The systems engineering process is well established and well understood. While this statement could be argued in the light of the many systems engineering guidelines and that have been developed, comparative review of these respective descriptions reveal that they differ primarily in the number of discrete steps or other nuances, and are at their core essentially common. Likewise, the systems engineering textbooks differ primarily in the context for application of systems engineering or in the utilization of evolved tools and techniques, not in the basic method. Thus, failures in systems engineering cannot credibly be attributed to implementation of the wrong systems engineering process among alternatives. However, numerous system failures can be attributed to deficient implementation of the systems engineering process. What may clearly be perceived as a systems engineering deficiency in retrospect can appear to be a well considered system engineering efficiency in real time—an efficiency taken to reduce cost or meet a schedule, or more often both. Typically these efficiencies are grounded on apparently solid rationale, such as reuse of heritage hardware or software. Over time, unintended consequences of a systems engineering process deficiency may begin to be realized, and unfortunately often the consequence is systems failure. This paper describes several actual cases of system failures that resulted from deficiencies in their systems engineering process implementation, including the Ariane 5 and the Hubble Space Telescope.

  2. The Therapy Process Observational Coding System for Child Psychotherapy Strategies Scale

    Science.gov (United States)

    McLeod, Bryce D.; Weisz, John R.

    2010-01-01

    Most everyday child and adolescent psychotherapy does not follow manuals that document the procedures. Consequently, usual clinical care has remained poorly understood and rarely studied. The Therapy Process Observational Coding System for Child Psychotherapy-Strategies scale (TPOCS-S) is an observational measure of youth psychotherapy procedures…

  3. 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)

    2004-03-01

    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)

  4. Toward a Systemic Notion of Information: Practical Consequences

    Directory of Open Access Journals (Sweden)

    Nagib Callaos

    2002-01-01

    Full Text Available Our main purpose in this paper is to start a process of a systemic definition of the notion of information and to provide some initial practical consequences of it. We will try to do that providing: 1 a conceptual definition, following Ackoff's (1962 description and method of such a kind of definition, and 2 following Peirce's (1931-5,1958 conception of "meaning", where the practical consequences should be included. To our knowledge, no attempt has been done up to the present neither to find a Peircean meaning to the notion of information, nor to start a process of describing a systemic notion of information. Consequently, we will try to integrate the different definitions made on information. But to integrate we should first differentiate what is to be integrated. Thus, we will typify information conceptions in subjective and objective, providing brief description and analysis of each type, integrating them in the context of a systemic notion of information, and drawing the respective pragmatic consequences, as required by Peirce, for any meaning description, and by a pragmatic-teleological systemic epistemology (Churchmann, 1971

  5. Modern Nuclear Data Evaluation with the TALYS Code System

    Science.gov (United States)

    Koning, A. J.; Rochman, D.

    2012-12-01

    This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: "Total" Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.

  6. Modern Nuclear Data Evaluation with the TALYS Code System

    International Nuclear Information System (INIS)

    Koning, A.J.; Rochman, D.

    2012-01-01

    This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: “Total” Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.

  7. Los Alamos radiation transport code system on desktop computing platforms

    International Nuclear Information System (INIS)

    Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.

    1990-01-01

    The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines

  8. Performance enhancement of optical code-division multiple-access systems using transposed modified Walsh code

    Science.gov (United States)

    Sikder, Somali; Ghosh, Shila

    2018-02-01

    This paper presents the construction of unipolar transposed modified Walsh code (TMWC) and analysis of its performance in optical code-division multiple-access (OCDMA) systems. Specifically, the signal-to-noise ratio, bit error rate (BER), cardinality, and spectral efficiency were investigated. The theoretical analysis demonstrated that the wavelength-hopping time-spreading system using TMWC was robust against multiple-access interference and more spectrally efficient than systems using other existing OCDMA codes. In particular, the spectral efficiency was calculated to be 1.0370 when TMWC of weight 3 was employed. The BER and eye pattern for the designed TMWC were also successfully obtained using OptiSystem simulation software. The results indicate that the proposed code design is promising for enhancing network capacity.

  9. Coded diffraction system in X-ray crystallography using a boolean phase coded aperture approximation

    Science.gov (United States)

    Pinilla, Samuel; Poveda, Juan; Arguello, Henry

    2018-03-01

    Phase retrieval is a problem present in many applications such as optics, astronomical imaging, computational biology and X-ray crystallography. Recent work has shown that the phase can be better recovered when the acquisition architecture includes a coded aperture, which modulates the signal before diffraction, such that the underlying signal is recovered from coded diffraction patterns. Moreover, this type of modulation effect, before the diffraction operation, can be obtained using a phase coded aperture, just after the sample under study. However, a practical implementation of a phase coded aperture in an X-ray application is not feasible, because it is computationally modeled as a matrix with complex entries which requires changing the phase of the diffracted beams. In fact, changing the phase implies finding a material that allows to deviate the direction of an X-ray beam, which can considerably increase the implementation costs. Hence, this paper describes a low cost coded X-ray diffraction system based on block-unblock coded apertures that enables phase reconstruction. The proposed system approximates the phase coded aperture with a block-unblock coded aperture by using the detour-phase method. Moreover, the SAXS/WAXS X-ray crystallography software was used to simulate the diffraction patterns of a real crystal structure called Rhombic Dodecahedron. Additionally, several simulations were carried out to analyze the performance of block-unblock approximations in recovering the phase, using the simulated diffraction patterns. Furthermore, the quality of the reconstructions was measured in terms of the Peak Signal to Noise Ratio (PSNR). Results show that the performance of the block-unblock phase coded apertures approximation decreases at most 12.5% compared with the phase coded apertures. Moreover, the quality of the reconstructions using the boolean approximations is up to 2.5 dB of PSNR less with respect to the phase coded aperture reconstructions.

  10. Noncoherent Spectral Optical CDMA System Using 1D Active Weight Two-Code Keying Codes

    Directory of Open Access Journals (Sweden)

    Bih-Chyun Yeh

    2016-01-01

    Full Text Available We propose a new family of one-dimensional (1D active weight two-code keying (TCK in spectral amplitude coding (SAC optical code division multiple access (OCDMA networks. We use encoding and decoding transfer functions to operate the 1D active weight TCK. The proposed structure includes an optical line terminal (OLT and optical network units (ONUs to produce the encoding and decoding codes of the proposed OLT and ONUs, respectively. The proposed ONU uses the modified cross-correlation to remove interferences from other simultaneous users, that is, the multiuser interference (MUI. When the phase-induced intensity noise (PIIN is the most important noise, the modified cross-correlation suppresses the PIIN. In the numerical results, we find that the bit error rate (BER for the proposed system using the 1D active weight TCK codes outperforms that for two other systems using the 1D M-Seq codes and 1D balanced incomplete block design (BIBD codes. The effective source power for the proposed system can achieve −10 dBm, which has less power than that for the other systems.

  11. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)

  12. Use of computer codes for system reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).

  13. 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.

  14. Vulnerability of ecological systems for nuclear war climatic consequences

    International Nuclear Information System (INIS)

    Kharuehll, M.; Khatchinson, T.; Kropper, U.; Kharuehll, K.

    1988-01-01

    Vulnerability of ecological systems of Northern hemisphere (terrestrial, aquatic and tropical) as well as Southern one in relation to climatic changes following large nuclear war is considered. When analyzing potential sensitivity of ecological systems to climatic changes, possible consequences are considered for different stress categories under various war scenarios. The above-mentioned stresses correspond to those adopted in published work by Pittok and others. To estimate the less important climatic disturbances a few additional computer-simulated models are developed

  15. Uncertainty and sensitivity analysis using probabilistic system assessment code. 1

    International Nuclear Information System (INIS)

    Honma, Toshimitsu; Sasahara, Takashi.

    1993-10-01

    This report presents the results obtained when applying the probabilistic system assessment code under development to the PSACOIN Level 0 intercomparison exercise organized by the Probabilistic System Assessment Code User Group in the Nuclear Energy Agency (NEA) of OECD. This exercise is one of a series designed to compare and verify probabilistic codes in the performance assessment of geological radioactive waste disposal facilities. The computations were performed using the Monte Carlo sampling code PREP and post-processor code USAMO. The submodels in the waste disposal system were described and coded with the specification of the exercise. Besides the results required for the exercise, further additional uncertainty and sensitivity analyses were performed and the details of these are also included. (author)

  16. 14 CFR Sec. 1-4 - System of accounts coding.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false System of accounts coding. Sec. 1-4 Section... General Accounting Provisions Sec. 1-4 System of accounts coding. (a) A four digit control number is assigned for each balance sheet and profit and loss account. Each balance sheet account is numbered...

  17. Performance Analysis of Optical Code Division Multiplex System

    Science.gov (United States)

    Kaur, Sandeep; Bhatia, Kamaljit Singh

    2013-12-01

    This paper presents the Pseudo-Orthogonal Code generator for Optical Code Division Multiple Access (OCDMA) system which helps to reduce the need of bandwidth expansion and improve spectral efficiency. In this paper we investigate the performance of multi-user OCDMA system to achieve data rate more than 1 Tbit/s.

  18. Data exchange between zero dimensional code and physics platform in the CFETR integrated system code

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)

    2016-11-01

    Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.

  19. The applicability of ALPHA/PHOENIX/ANC nuclear design code system on Korean standard PWR's

    International Nuclear Information System (INIS)

    Lee, Kookjong; Choi, Kie-Yong; Lee, Hae-Chan; Roh, Eun-Rae

    1996-01-01

    For the Korean Standard Nuclear Power Plant (KSNPP) designed based on Combustion Engineering (CE) System 80, the Westinghouse nuclear design code system ALPHA/PHOENIX/ANC was applied to the follow-up design of initial and reload core of KSNPP. The follow-up design results of Yonggwang Unit 3 Cycle 1, 2 and Yonggwang Unit 4 Cycle 1 have shown good agreements with the measured data. The assemblywise power distributions have shown less than 2% average differences and critical boron concentrations have shown less than 20 ppm differences. All the low power physics test parameters are in good agreement. Consequently, APA design code system can be applied to KNSPP cores. (author)

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

    International Nuclear Information System (INIS)

    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

  1. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  2. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I

    1985-07-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs.

  3. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs

  4. SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi

    2003-01-01

    SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)

  5. Performance optimization of spectral amplitude coding OCDMA system using new enhanced multi diagonal code

    Science.gov (United States)

    Imtiaz, Waqas A.; Ilyas, M.; Khan, Yousaf

    2016-11-01

    This paper propose a new code to optimize the performance of spectral amplitude coding-optical code division multiple access (SAC-OCDMA) system. The unique two-matrix structure of the proposed enhanced multi diagonal (EMD) code and effective correlation properties, between intended and interfering subscribers, significantly elevates the performance of SAC-OCDMA system by negating multiple access interference (MAI) and associated phase induce intensity noise (PIIN). Performance of SAC-OCDMA system based on the proposed code is thoroughly analyzed for two detection techniques through analytic and simulation analysis by referring to bit error rate (BER), signal to noise ratio (SNR) and eye patterns at the receiving end. It is shown that EMD code while using SDD technique provides high transmission capacity, reduces the receiver complexity, and provides better performance as compared to complementary subtraction detection (CSD) technique. Furthermore, analysis shows that, for a minimum acceptable BER of 10-9 , the proposed system supports 64 subscribers at data rates of up to 2 Gbps for both up-down link transmission.

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

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    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)

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

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

    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

  8. 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.

  9. The radiological assessment system for consequence analysis - RASCAL

    International Nuclear Information System (INIS)

    Sjoreen, A.L.; Ramsdell, J.V.; Athey, G.F.

    1996-01-01

    The Radiological Assessment System for Consequence Analysis, Version 2.1 (RASCAL 2.1) has been developed for use during a response to radiological emergencies. The model estimates doses for comparison with U.S. Environmental Protection Agency (EPA) Protective Action Guides (PAGs) and thresholds for acute health effects. RASCAL was designed to be used by U.S. Nuclear Regulatory Commission (NRC) personnel who report to the site of a nuclear accident to conduct an independent evaluation of dose and consequence projections and personnel who conduct training and drills on emergency responses. It allows consideration of the dominant aspects of the source term, transport, dose, and consequences. RASCAL consists of three computational tools: ST-DOSE, FM-DOSE, and DECAY. ST-DOSE computes source term, atmospheric transport, and dose to man from accidental airborne releases of radionuclides. The source-term calculations are appropriate for accidents at U.S. power reactors. FM-DOSE computes doses from environmental concentrations of radionuclides in the air and on the ground. DECAY computes radiological decay and daughter in-growth. RASCAL 2.1 is a DOS application that can be run under Windows 3.1 and 95. RASCAL has been the starting point for other accident consequence models, notably INTERRAS, an international version of RASCAL, and HASCAL, an expansion of RASCAL that will model radiological, biological, and chemical accidents

  10. Locally Minimum Storage Regenerating Codes in Distributed Cloud Storage Systems

    Institute of Scientific and Technical Information of China (English)

    Jing Wang; Wei Luo; Wei Liang; Xiangyang Liu; Xiaodai Dong

    2017-01-01

    In distributed cloud storage sys-tems, inevitably there exist multiple node fail-ures at the same time. The existing methods of regenerating codes, including minimum storage regenerating (MSR) codes and mini-mum bandwidth regenerating (MBR) codes, are mainly to repair one single or several failed nodes, unable to meet the repair need of distributed cloud storage systems. In this paper, we present locally minimum storage re-generating (LMSR) codes to recover multiple failed nodes at the same time. Specifically, the nodes in distributed cloud storage systems are divided into multiple local groups, and in each local group (4, 2) or (5, 3) MSR codes are constructed. Moreover, the grouping method of storage nodes and the repairing process of failed nodes in local groups are studied. The-oretical analysis shows that LMSR codes can achieve the same storage overhead as MSR codes. Furthermore, we verify by means of simulation that, compared with MSR codes, LMSR codes can reduce the repair bandwidth and disk I/O overhead effectively.

  11. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Ryu, Chang Mo; Kim, Yeon Seung; Eom, Heung Seop; Lee, Jong Bok; Kim, Ho Joon; Choi, Young Gil; Kim, Ko Ryeo

    1989-01-01

    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)

  12. SURE: a system of computer codes for performing sensitivity/uncertainty analyses with the RELAP code

    International Nuclear Information System (INIS)

    Bjerke, M.A.

    1983-02-01

    A package of computer codes has been developed to perform a nonlinear uncertainty analysis on transient thermal-hydraulic systems which are modeled with the RELAP computer code. Using an uncertainty around the analyses of experiments in the PWR-BDHT Separate Effects Program at Oak Ridge National Laboratory. The use of FORTRAN programs running interactively on the PDP-10 computer has made the system very easy to use and provided great flexibility in the choice of processing paths. Several experiments simulating a loss-of-coolant accident in a nuclear reactor have been successfully analyzed. It has been shown that the system can be automated easily to further simplify its use and that the conversion of the entire system to a base code other than RELAP is possible

  13. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

    Ryman, J.C.; Cristy, M.; Eckerman, K.F.; Davis, J.L.; Tang, J.S.; Kerr, G.D.

    1986-01-01

    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

  14. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

    Han, S.H.; Yoo, K.J.; Kim, T.W.

    1987-01-01

    The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

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

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    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)

  16. A computerized energy systems code and information library at Soreq

    Energy Technology Data Exchange (ETDEWEB)

    Silverman, I; Shapira, M; Caner, D; Sapier, D [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

    1996-12-01

    In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors).

  17. A computerized energy systems code and information library at Soreq

    International Nuclear Information System (INIS)

    Silverman, I.; Shapira, M.; Caner, D.; Sapier, D.

    1996-01-01

    In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors)

  18. 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.

  19. 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...

  20. The PASC-3 code system and the UNIPASC environment

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Oppe, J.; Oudshoorn, H.

    1991-08-01

    A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and its associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified, Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab

  1. Sequence Coding and Search System Backfit Quality Assurance Program Plan

    International Nuclear Information System (INIS)

    Lovell, C.J.; Stepina, P.L.

    1985-03-01

    The Sequence Coding and Search System is a computer-based encoding system for events described in Licensee Event Reports. This data system contains LERs from 1981 to present. Backfit of the data system to include LERs prior to 1981 is required. This report documents the Quality Assurance Program Plan that EG and G Idaho, Inc. will follow while encoding 1980 LERs

  2. 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.

  3. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

    1997-07-01

    During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

  4. LHC beam dump system Consequences of abnormal operation

    CERN Document Server

    Kramer, T; Uythoven, J

    2010-01-01

    The LHC beam dump system is one of the most critical systems concerning machine protection and safe operation. It is used to dispose of high intensity beams between 450 GeV and 7 TeV. Studies into the consequences of abnormal beam dump actions have been performed. Different error scenarios have been evaluated using particle tracking in MAD-X, including an asynchronous dump action, and the impact of different orbit and collimator settings. Losses at locations in the ring and the beam dump transfer lines have been quantified as a function of different settings of the dump system protection elements. The implications for the setting up and operation of these protection elements are discussed.

  5. Dynamic detection technology of malicious code for Android system

    Directory of Open Access Journals (Sweden)

    Li Boya

    2017-02-01

    Full Text Available With the increasing popularization of mobile phones,people's dependence on them is rising,the security problems become more and more prominent.According to the calling of the APK file permission and the API function in Android system,this paper proposes a dynamic detecting method based on API interception technology to detect the malicious code.The experimental results show that this method can effectively detect the malicious code in Android system.

  6. Delay Estimation in Long-Code Asynchronous DS/CDMA Systems Using Multiple Antennas

    Directory of Open Access Journals (Sweden)

    Sirbu Marius

    2004-01-01

    Full Text Available The problem of propagation delay estimation in asynchronous long-code DS-CDMA multiuser systems is addressed. Almost all the methods proposed so far in the literature for propagation delay estimation are derived for short codes and the knowledge of the codes is exploited by the estimators. In long-code CDMA, the spreading code is aperiodic and the methods developed for short codes may not be used or may increase the complexity significantly. For example, in the subspace-based estimators, the aperiodic nature of the code may require subspace tracking. In this paper we propose a novel method for simultaneous estimation of the propagation delays of several active users. A specific multiple-input multiple-output (MIMO system model is constructed in a multiuser scenario. In such model the channel matrix contains information about both the users propagation delays and channel impulse responses. Consequently, estimates of the delays are obtained as a by-product of the channel estimation task. The channel matrix has a special structure that is exploited in estimating the delays. The proposed delay estimation method lends itself to an adaptive implementation. Thus, it may be applied to joint channel and delay estimation in uplink DS-CDMA analogously to the method presented by the authors in 2003. The performance of the proposed method is studied in simulation using realistic time-varying channel model and different SNR levels in the face of near-far effects, and using low spreading factor (high data rates.

  7. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

  8. Nonterminals and codings in defining variations of OL-systems

    DEFF Research Database (Denmark)

    Skyum, Sven

    1974-01-01

    The use of nonterminals versus the use of codings in variations of OL-systems is studied. It is shown that the use of nonterminals produces a comparatively low generative capacity in deterministic systems while it produces a comparatively high generative capacity in nondeterministic systems. Fina....... Finally it is proved that the family of context-free languages is contained in the family generated by codings on propagating OL-systems with a finite set of axioms, which was one of the open problems in [10]. All the results in this paper can be found in [71] and [72].......The use of nonterminals versus the use of codings in variations of OL-systems is studied. It is shown that the use of nonterminals produces a comparatively low generative capacity in deterministic systems while it produces a comparatively high generative capacity in nondeterministic systems...

  9. ATHENA code manual. Volume 1. Code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Carlson, K.E.; Roth, P.A.; Ransom, V.H.

    1986-09-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation

  10. Lossless Coding Standards for Space Data Systems

    Science.gov (United States)

    Rice, R. F.

    1996-01-01

    The International Consultative Committee for Space Data Systems (CCSDS) is preparing to issue its first recommendation for a digital data compression standard. Because the space data systems of primary interest are employed to support scientific investigations requiring accurate representation, this initial standard will be restricted to lossless compression.

  11. A code system for ADS transmutation studies

    International Nuclear Information System (INIS)

    Brolly, A.; Vertes, P.

    2001-01-01

    An accelerator driven reactor physical system can be divided into two different subsystems. One is the neutron source the other is the subcritical reactor. Similarly, the modelling of such system is also split into two parts. The first step is the determination of the spatial distribution and angle-energy spectrum of neutron source in the target region; the second one is the calculation of neutron flux which is responsible for the transmutation process in the subcritical system. Accelerators can make neutrons from high energy protons by spallation or photoneutrons from accelerated electrons by Bremsstrahlung (e-n converter). The Monte Carlo approach is the only way of modelling such processes and it might be extended to the whole subcritical system as well. However, a subcritical reactor may be large, it may contain thermal regions and the lifetime of neutrons may be long. Therefore a comprehensive Monte Carlo modelling of such system is a very time consuming computational process. It is unprofitable as well when applied to system optimization that requires a comparative study of large number of system variants. An appropriate method of deterministic transport calculation may adequately satisfy these requirements. Thus, we have built up a coupled calculational model for ADS to be used for transmutation of nuclear waste which we refer further as M-c-T system. Flow chart is shown in Figure. (author)

  12. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  13. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

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

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    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

  15. Automatic code generation for distributed robotic systems

    International Nuclear Information System (INIS)

    Jones, J.P.

    1993-01-01

    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

  16. Hydrogen detection systems leak response codes

    International Nuclear Information System (INIS)

    Desmas, T.; Kong, N.; Maupre, J.P.; Schindler, P.; Blanc, D.

    1990-01-01

    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

  17. Source Code Vulnerabilities in IoT Software Systems

    Directory of Open Access Journals (Sweden)

    Saleh Mohamed Alnaeli

    2017-08-01

    Full Text Available An empirical study that examines the usage of known vulnerable statements in software systems developed in C/C++ and used for IoT is presented. The study is conducted on 18 open source systems comprised of millions of lines of code and containing thousands of files. Static analysis methods are applied to each system to determine the number of unsafe commands (e.g., strcpy, strcmp, and strlen that are well-known among research communities to cause potential risks and security concerns, thereby decreasing a system’s robustness and quality. These unsafe statements are banned by many companies (e.g., Microsoft. The use of these commands should be avoided from the start when writing code and should be removed from legacy code over time as recommended by new C/C++ language standards. Each system is analyzed and the distribution of the known unsafe commands is presented. Historical trends in the usage of the unsafe commands of 7 of the systems are presented to show how the studied systems evolved over time with respect to the vulnerable code. The results show that the most prevalent unsafe command used for most systems is memcpy, followed by strlen. These results can be used to help train software developers on secure coding practices so that they can write higher quality software systems.

  18. 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

  19. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  20. System verification and validation report for the TMAD code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the Verification and Validation Report for the TMAD code system, which includes the TMAD code and the LIBMAKR Code. The TMAD code was commissioned to facilitate the interpretation of moisture probe measurements in the Hanford Site waste tanks. In principle, the code is an interpolation routine that acts over a library of benchmark data based on two independent variables, typically anomaly size and moisture content. Two additional variables, anomaly type and detector type, can also be considered independent variables, but no interpolation is done over them. The dependent variable is detector response. The intent is to provide the code with measured detector responses from two or more detectors. The code will then interrogate (and interpolate upon) the benchmark data library and find the anomaly-type/anomaly-size/moisture-content combination that provides the closest match to the measured data. The primary purpose of this document is to provide the results of the system testing and the conclusions based thereon. The results of the testing process are documented in the body of the report. Appendix A gives the test plan, including test procedures, used in conducting the tests. Appendix B lists the input data required to conduct the tests, and Appendices C and 0 list the numerical results of the tests

  1. Integrated Validation System for a Thermal-hydraulic System Code, TASS/SMR-S

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee-Kyung; Kim, Hyungjun; Kim, Soo Hyoung; Hwang, Young-Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Hyeon-Soo [Chungnam National University, Daejeon (Korea, Republic of)

    2015-10-15

    Development including enhancement and modification of thermal-hydraulic system computer code is indispensable to a new reactor, SMART. Usually, a thermal-hydraulic system code validation is achieved by a comparison with the results of corresponding physical effect tests. In the reactor safety field, a similar concept, referred to as separate effect tests has been used for a long time. But there are so many test data for comparison because a lot of separate effect tests and integral effect tests are required for a code validation. It is not easy to a code developer to validate a computer code whenever a code modification is occurred. IVS produces graphs which shown the comparison the code calculation results with the corresponding test results automatically. IVS was developed for a validation of TASS/SMR-S code. The code validation could be achieved by a comparison code calculation results with corresponding test results. This comparison was represented as a graph for convenience. IVS is useful before release a new code version. The code developer can validate code result easily using IVS. Even during code development, IVS could be used for validation of code modification. The code developer could gain a confidence about his code modification easily and fast and could be free from tedious and long validation work. The popular software introduced in IVS supplies better usability and portability.

  2. 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.

  3. Establishment of computer code system for nuclear reactor design - analysis

    International Nuclear Information System (INIS)

    Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.

    1996-01-01

    Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab

  4. Grid-code of Croatian power system

    International Nuclear Information System (INIS)

    Toljan, I.; Mesic, M.; Kalea, M.; Koscak, Z.

    2003-01-01

    Grid Rules by the Croatian Electricity Utility deal with the control and usage of the Croatian power system's transmission and distribution grid. Furthermore, these rules include obligations and permissions of power grid users and owners, with the aim of a reliable electricity supply.(author)

  5. Summary description of the scale modular code system

    International Nuclear Information System (INIS)

    Parks, C.V.

    1987-12-01

    SCALE - a modular code system for Standardized Computer Analyses for Licensing Evaluation - has been developed at Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission staff. The SCALE system utilizes well-established computer codes and methods within standard analytic sequences that allow simplified free-form input, automate the data processing and coupling between codes, and provide accurate and reliable results. System development has been directed at criticality safety, shielding, and heat transfer analysis of spent fuel transport and/or storage casks. However, only a few of the sequences (and none of the individual functional modules) are restricted to cask applications. This report will provide a background on the history of the SCALE development and review the components and their function within the system. The available data libraries are also discussed, together with the automated features that standardize the data processing and systems analysis. 83 refs., 32 figs., 11 tabs

  6. On the automated assessment of nuclear reactor systems code accuracy

    International Nuclear Information System (INIS)

    Kunz, Robert F.; Kasmala, Gerald F.; Mahaffy, John H.; Murray, Christopher J.

    2002-01-01

    An automated code assessment program (ACAP) has been developed to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. The tool provides a suite of metrics for quality of fit to specific data sets, and the means to produce one or more figures of merit (FOM) for a code, based on weighted averages of results from the batch execution of a large number of code-experiment and code-code data comparisons. Accordingly, this tool has the potential to significantly streamline the verification and validation (V and V) processes in NRS code development environments which are characterized by rapidly evolving software, many contributing developers and a large and growing body of validation data. In this paper, a survey of data conditioning and analysis techniques is summarized which focuses on their relevance to NRS code accuracy assessment. A number of methods are considered for their applicability to the automated assessment of the accuracy of NRS code simulations. A variety of data types and computational modeling methods are considered from a spectrum of mathematical and engineering disciplines. The goal of the survey was to identify needs, issues and techniques to be considered in the development of an automated code assessment procedure, to be used in United States Nuclear Regulatory Commission (NRC) advanced thermal-hydraulic T/H code consolidation efforts. The ACAP software was designed based in large measure on the findings of this survey. An overview of this tool is summarized and several NRS data applications are provided. The paper is organized as follows: The motivation for this work is first provided by background discussion that summarizes the relevance of this subject matter to the nuclear reactor industry. Next, the spectrum of NRS data types are classified into categories, in order to provide a basis for assessing individual comparison methods. Then, a summary of the survey is provided, where each

  7. The program system UFOMOD for assessing the consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Burkart, K.; Hasemann, I.; Matzerath, C.; Panitz, H.J.; Steinhauer, C.

    1988-10-01

    The programm system UFOMOD is a completely new accident consequence assessment (ACA) code. Its structure and modelling is based on the experience gained from applications of the old UFOMOD code during and after the German Risk Study - Phase A, the results of scientific investigations performed within the ongoing Phase B and the CEC-project MARIA, and the requirements resulting from the extended use of ACAs to help in decision-making. One of the most important improvements is the introduction of different trajecotry models for describing atmospheric dispersion in the near range and at larger distances. Emergency actions and countermeasures modelling takes into account recommendations of international commissions. The dosimetric models contain completely new age-, sex- and time-dependent data of dose-conversion factors for external and internal radiation; the ingestion pathway is modelled to consider seasonal dependencies. New dose-risk-relationships for stochastic and non-stochastic health effects are implemented; a special algorithm developed for ACA codes allows individual and collective leukemia and cancer risks to be presented as a function of time after the accident. According to the modular structure of the new program system UFOMOD, an easy access to parameter values and the results of the various submodels exists what facilitates sensitivity and uncertainty analyses. (orig.) [de

  8. Analytical considerations in the code qualification of piping systems

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1995-01-01

    The paper addresses several analytical topics in the design and qualification of piping systems which have a direct bearing on the prediction of stresses in the pipe and hence on the application of the equations of NB, NC and ND-3600 of the ASME Boiler and Pressure Vessel Code. For each of the analytical topics, the paper summarizes the current code requirements, if any, and the industry practice

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

    International Nuclear Information System (INIS)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C.; Galloway, Jack

    2011-01-01

    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

  10. 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

  11. Transient and fuel performance analysis with VTT's coupled code system

    International Nuclear Information System (INIS)

    Daavittila, A.; Hamalainen, A.; Raty, H.

    2005-01-01

    VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

  12. ARAC: A flexible real-time dose consequence assessment system

    International Nuclear Information System (INIS)

    Ellis, J.S.; Sullivan, T.J.

    1993-01-01

    Since its beginning, the Atmospheric Release Advisory Capability (ARAC), an emergency radiological dose assessment service of the US Government, has been called on to do consequence assessments for releases into the atmosphere of radionuclides and a variety of other substances. Some of the more noteworthy emergency responses have been for the Three Mile Island and Chernobyl nuclear power reactor accidents, and more recently, for a cloud of gases from a rail-car spill into the Sacramento river of the herbicide metam sodium, smoke from hundreds of burning oil wells in Kuwait, and ash clouds from the eruption of Mt. Pinatubo. The spatial scales of these responses range from local, to regional, to global, and the response periods from hours, to weeks, to months. Because of the variety of requirements of each unique assessment, ARAC has developed and maintains a flexible system of people, computer software and hardware

  13. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. H.; Kim, T. K.; Kang, I. S.; Cho, H. S.; Son, J. S. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    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.

  15. Improvement of JRR-4 core management code system

    International Nuclear Information System (INIS)

    Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

    2000-01-01

    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)

  16. Development of the vacuum system pressure responce analysis code PRAC

    International Nuclear Information System (INIS)

    Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.

    1985-03-01

    In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)

  17. Bilingual processing of ASL-English code-blends: The consequences of accessing two lexical representations simultaneously

    OpenAIRE

    Emmorey, Karen; Petrich, Jennifer; Gollan, Tamar H.

    2012-01-01

    Bilinguals who are fluent in American Sign Language (ASL) and English often produce code-blends - simultaneously articulating a sign and a word while conversing with other ASL-English bilinguals. To investigate the cognitive mechanisms underlying code-blend processing, we compared picture-naming times (Experiment 1) and semantic categorization times (Experiment 2) for code-blends versus ASL signs and English words produced alone. In production, code-blending did not slow lexical retrieval for...

  18. Support of Maritime Education and Training Systems for the Implementation of the ISM Code

    Directory of Open Access Journals (Sweden)

    Elena Maggi

    2012-10-01

    Full Text Available The problem of the safety improvement and pollution preventionin maritime tra11Sp011 has become more and more criticaland urgent to solve. In fact, the number of accidents at seahas increased very quickly over time. Concern is growing aboutpoor qualification on safety and poor management standardsin shipping industry. The ISM Code (Intenzational Safety ManagementCode aims to provide intemational standards for safemanagement of ship operations and for pollution prevention.The paper, which presents a part of the work done by the Universityof Trieste - ISTIEE within the framework of theMETHAR project, aims to identify the expected supp01t fromthe MaJitime Education and Training (MET systems to implemmttheISM Code, consequently improving safety and preventingpollution. The paper desCiibes first the 01igin, the objectivesof the ISM Code and the standard requirements on METidentified by the Code. Secondly, it summarises the opinions ofthe operators collected through questionnaires. Finally, it identifiesthe possible enrichment of the MET system in order tobel/.er optimise the implementation of the Code.

  19. Subchannel analysis of a boiloff experiment by a system thermalhydraulic code

    International Nuclear Information System (INIS)

    Bousbia-Salah, A.; D'Auria, F.

    2001-01-01

    This paper presents the results of system thermalhydraulic code using the sub-channel analysis approach in predicting the Neptun boil off experiments. This approach will be suitable for further works in view of coupling the system code with a 3D neutron kinetic one. 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 No5002, which is a good repeat experiment, is considered. The calculations were carried out using the system transient analysis code Relap5/Mod3.2. A detailed nodalization of the Neptun test section was developed. 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, this demonstrates, as well, the reasonable success of the subchannel analysis approach adopted in the present context for a system thermalhydraulic code.(author)

  20. 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.

  1. Neurological consequences of systemic inflammation in the premature neonate.

    Science.gov (United States)

    Patra, Aparna; Huang, Hong; Bauer, John A; Giannone, Peter J

    2017-06-01

    Despite substantial progress in neonatal care over the past two decades leading to improved survival of extremely premature infants, extreme prematurity continues to be associated with long term neurodevelopmental impairments. Cerebral white matter injury is the predominant form of insult in preterm brain leading to adverse neurological consequences. Such brain injury pattern and unfavorable neurologic sequelae is commonly encountered in premature infants exposed to systemic inflammatory states such as clinical or culture proven sepsis with or without evidence of meningitis, prolonged mechanical ventilation, bronchopulmonary dysplasia, necrotizing enterocolitis and chorioamnionitis. Underlying mechanisms may include cytokine mediated processes without direct entry of pathogens into the brain, developmental differences in immune response and complex neurovascular barrier system that play a critical role in regulating the cerebral response to various systemic inflammatory insults in premature infants. Understanding of these pathologic mechanisms and clinical correlates of such injury based on serum biomarkers or brain imaging findings on magnetic resonance imaging will pave way for future research and translational therapeutic opportunities for the developing brain.

  2. Neurological consequences of systemic inflammation in the premature neonate

    Directory of Open Access Journals (Sweden)

    Aparna Patra

    2017-01-01

    Full Text Available Despite substantial progress in neonatal care over the past two decades leading to improved survival of extremely premature infants, extreme prematurity continues to be associated with long term neurodevelopmental impairments. Cerebral white matter injury is the predominant form of insult in preterm brain leading to adverse neurological consequences. Such brain injury pattern and unfavorable neurologic sequelae is commonly encountered in premature infants exposed to systemic inflammatory states such as clinical or culture proven sepsis with or without evidence of meningitis, prolonged mechanical ventilation, bronchopulmonary dysplasia, necrotizing enterocolitis and chorioamnionitis. Underlying mechanisms may include cytokine mediated processes without direct entry of pathogens into the brain, developmental differences in immune response and complex neurovascular barrier system that play a critical role in regulating the cerebral response to various systemic inflammatory insults in premature infants. Understanding of these pathologic mechanisms and clinical correlates of such injury based on serum biomarkers or brain imaging findings on magnetic resonance imaging will pave way for future research and translational therapeutic opportunities for the developing brain.

  3. A Coding System for Analysing a Spoken Text Database.

    Science.gov (United States)

    Cutting, Joan

    1994-01-01

    This paper describes a coding system devised to analyze conversations of graduate students in applied linguistics at Edinburgh University. The system was devised to test the hypothesis that as shared knowledge among conversation participants grows, the textual density of in-group members has more cues than that of strangers. The informal…

  4. Physical-layer network coding in coherent optical OFDM systems.

    Science.gov (United States)

    Guan, Xun; Chan, Chun-Kit

    2015-04-20

    We present the first experimental demonstration and characterization of the application of optical physical-layer network coding in coherent optical OFDM systems. It combines two optical OFDM frames to share the same link so as to enhance system throughput, while individual OFDM frames can be recovered with digital signal processing at the destined node.

  5. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

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

    International Nuclear Information System (INIS)

    Liu, X.J.; Yang, T.; Cheng, X.

    2014-01-01

    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)

  7. Progress on China nuclear data processing code system

    Science.gov (United States)

    Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

    2017-09-01

    China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

  8. Code conversion for system design and safety analysis of NSSS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)

  9. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  10. Environmental consequences of different beef production systems in the EU

    DEFF Research Database (Denmark)

    Nguyen, T Lan T; Hermansen, John Erik; Mogensen, Lisbeth

    2010-01-01

    The aim of this paper is to examine the environmental consequences of beef meat production in the EU, using a life cycle approach. Four beef production systems were studied - three from intensively reared dairy calves and one from suckler herds. According to the results of the analysis......, the contributions from the production of 1 kg beef meat (slaughter weight) to global warming, acidification, eutrophication, land use and non-renewable energy use were lower for beef from dairy calves than from suckler herds (16.0-19.9 versus 27.3 kg CO2e, 101-173 versus 210 g SO2e, 622-1140 versus 1651 g NO3e, 16.......5-22.7 versus 42.9 m2year, and 41.3-48.2 versus 59.2 MJ, respectively). The breakdown analysis helped identify the key areas in the "cradle to farm gate" beef production system where sustainable management strategies are needed to improve environmental performance. The study also included a sensitivity analysis...

  11. Solar System Chaos and its climatic and biogeochemical consequences

    Science.gov (United States)

    Ikeda, M.; Tada, R.; Ozaki, K.; Olsen, P. E.

    2017-12-01

    Insolation changes caused by changes in Earth's orbital parameters are the main driver of climatic variations, whose pace has been used for astronomically-calibrated geologic time scales of high accuracy to understand Earth system dynamics. However, the astrophysical models beyond several tens of million years ago have large uncertainty due to chaotic behavior of the Solar System, and its impact on amplitude modulation of multi-Myr-scale orbital variations and consequent climate changes has become the subject of debate. Here we show the geologic constraints on the past chaotic behavior of orbital cycles from early Mesozoic monsoon-related records; the 30-Myr-long lake level records of the lacustrine sequence in Newark-Hartford basins (North America) and 70-Myr-long biogenic silica (BSi) burial flux record of pelagic deep-sea chert sequence in Inuyama area (Japan). BSi burial flux of chert could be considered as proportional to the dissolved Si (DSi) input from chemical weathering on timescales longer than the residence time of DSi ( 100 kyr), because chert could represent a major sink for oceanic dissolved silica (Ikeda et al., 2017).These geologic records show multi-Myr cycles with similar frequency modulations of eccentricity solution of astronomical model La2010d (Laskar et al., 2011) compared with other astronomical solutions, but not exactly same. Our geologic records provide convincing evidence for the past chaotic dynamical behaviour of the Solar System and new and challenging additional constraints for astrophysical models. In addition, we find that ˜10 Myr cycle detected in monsoon proxies and their amplitude modulation of ˜2 Myr cycle may be related to the amplitude modulation of ˜2 Myr eccentricity cycle through non-linear process(es) of Earth system dynamics, suggesting possible impact of the chaotic behavior of Solar planets on climate change. Further impact of multi-Myr orbital cycles on global biogeochemical cycles will be discussed.

  12. Simulation of water hammer phenomena using the system code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-15

    Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

  13. Simulation of water hammer phenomena using the system code ATHLET

    International Nuclear Information System (INIS)

    Bratfisch, Christoph; Koch, Marco K.

    2017-01-01

    Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

  14. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  15. Development of FBR integrity system code. Basic concept

    International Nuclear Information System (INIS)

    Asayama, Tai

    2001-05-01

    For fast breeder reactors to be commercialized, they must be more reliable, safer, and at the same, economically competitive with future light water reactors. Innovation of elevated temperature structural design standard is necessary to achieve this goal. The most powerful way is to enlarge the scope of structural integrity code to cover items other than design evaluation that has been addressed in existing codes. Items that must be newly covered are prerequisites of design, fabrication, examination, operation and maintenance, etc. This allows designers to choose the most economical combination of design variations to achieve specific reliability that is needed for a particular component. Designing components by this concept, a cost-minimum design of a whole plant can be realized. By determining the reliability that must be achieved for a component by risk technologies, further economical improvement can be expected by avoiding excessive quality. Recognizing the necessity for the codes based on the new concept, the development of 'FBR integrity system code' began in 2000. Research and development will last 10 years. For this development, the basic logistics and system as well as technologies that materialize the concept are necessary. Original logistics and system must be developed, because no existing researches are available in and out of Japan. This reports presents the results of the work done in the first year regarding the basic idea, methodology, and structure of the code. (author)

  16. Java Source Code Analysis for API Migration to Embedded Systems

    Energy Technology Data Exchange (ETDEWEB)

    Winter, Victor [Univ. of Nebraska, Omaha, NE (United States); McCoy, James A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Guerrero, Jonathan [Univ. of Nebraska, Omaha, NE (United States); Reinke, Carl Werner [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perry, James Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-01

    Embedded systems form an integral part of our technological infrastructure and oftentimes play a complex and critical role within larger systems. From the perspective of reliability, security, and safety, strong arguments can be made favoring the use of Java over C in such systems. In part, this argument is based on the assumption that suitable subsets of Java’s APIs and extension libraries are available to embedded software developers. In practice, a number of Java-based embedded processors do not support the full features of the JVM. For such processors, source code migration is a mechanism by which key abstractions offered by APIs and extension libraries can made available to embedded software developers. The analysis required for Java source code-level library migration is based on the ability to correctly resolve element references to their corresponding element declarations. A key challenge in this setting is how to perform analysis for incomplete source-code bases (e.g., subsets of libraries) from which types and packages have been omitted. This article formalizes an approach that can be used to extend code bases targeted for migration in such a manner that the threats associated the analysis of incomplete code bases are eliminated.

  17. Differential Space-Time Block Code Modulation for DS-CDMA Systems

    Directory of Open Access Journals (Sweden)

    Liu Jianhua

    2002-01-01

    Full Text Available A differential space-time block code (DSTBC modulation scheme is used to improve the performance of DS-CDMA systems in fast time-dispersive fading channels. The resulting scheme is referred to as the differential space-time block code modulation for DS-CDMA (DSTBC-CDMA systems. The new modulation and demodulation schemes are especially studied for the down-link transmission of DS-CDMA systems. We present three demodulation schemes, referred to as the differential space-time block code Rake (D-Rake receiver, differential space-time block code deterministic (D-Det receiver, and differential space-time block code deterministic de-prefix (D-Det-DP receiver, respectively. The D-Det receiver exploits the known information of the spreading sequences and their delayed paths deterministically besides the Rake type combination; consequently, it can outperform the D-Rake receiver, which employs the Rake type combination only. The D-Det-DP receiver avoids the effect of intersymbol interference and hence can offer better performance than the D-Det receiver.

  18. Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

    International Nuclear Information System (INIS)

    Goto, Minoru; Takamatsu, Kuniyoshi

    2007-03-01

    The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)

  19. Opacity calculations for extreme physical systems: code RACHEL

    Science.gov (United States)

    Drska, Ladislav; Sinor, Milan

    1996-08-01

    Computer simulations of physical systems under extreme conditions (high density, temperature, etc.) require the availability of extensive sets of atomic data. This paper presents basic information on a self-consistent approach to calculations of radiative opacity, one of the key characteristics of such systems. After a short explanation of general concepts of the atomic physics of extreme systems, the structure of the opacity code RACHEL is discussed and some of its applications are presented.

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

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Sanchez Espinoza, Victor H.; Feng, Bo

    2008-01-01

    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)

  1. 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)

  2. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    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)

  3. Coded aperture imaging system for nuclear fuel motion detection

    International Nuclear Information System (INIS)

    Stalker, K.T.; Kelly, J.G.

    1980-01-01

    A Coded Aperature Imaging System (CAIS) has been developed at Sandia National Laboratories to image the motion of nuclear fuel rods undergoing tests simulating accident conditions within a liquid metal fast breeder reactor. The tests require that the motion of the test fuel be monitored while it is immersed in a liquid sodium coolant precluding the use of normal optical means of imaging. However, using the fission gamma rays emitted by the fuel itself and coded aperture techniques, images with 1.5 mm radial and 5 mm axial resolution have been attained. Using an electro-optical detection system coupled to a high speed motion picture camera a time resolution of one millisecond can be achieved. This paper will discuss the application of coded aperture imaging to the problem, including the design of the one-dimensional Fresnel zone plate apertures used and the special problems arising from the reactor environment and use of high energy gamma ray photons to form the coded image. Also to be discussed will be the reconstruction techniques employed and the effect of various noise sources on system performance. Finally, some experimental results obtained using the system will be presented

  4. Adaptive Wavelet Coding Applied in a Wireless Control System.

    Science.gov (United States)

    Gama, Felipe O S; Silveira, Luiz F Q; Salazar, Andrés O

    2017-12-13

    Wireless control systems can sense, control and act on the information exchanged between the wireless sensor nodes in a control loop. However, the exchanged information becomes susceptible to the degenerative effects produced by the multipath propagation. In order to minimize the destructive effects characteristic of wireless channels, several techniques have been investigated recently. Among them, wavelet coding is a good alternative for wireless communications for its robustness to the effects of multipath and its low computational complexity. This work proposes an adaptive wavelet coding whose parameters of code rate and signal constellation can vary according to the fading level and evaluates the use of this transmission system in a control loop implemented by wireless sensor nodes. The performance of the adaptive system was evaluated in terms of bit error rate (BER) versus E b / N 0 and spectral efficiency, considering a time-varying channel with flat Rayleigh fading, and in terms of processing overhead on a control system with wireless communication. The results obtained through computational simulations and experimental tests show performance gains obtained by insertion of the adaptive wavelet coding in a control loop with nodes interconnected by wireless link. These results enable the use of this technique in a wireless link control loop.

  5. Adaptive Wavelet Coding Applied in a Wireless Control System

    Directory of Open Access Journals (Sweden)

    Felipe O. S. Gama

    2017-12-01

    Full Text Available Wireless control systems can sense, control and act on the information exchanged between the wireless sensor nodes in a control loop. However, the exchanged information becomes susceptible to the degenerative effects produced by the multipath propagation. In order to minimize the destructive effects characteristic of wireless channels, several techniques have been investigated recently. Among them, wavelet coding is a good alternative for wireless communications for its robustness to the effects of multipath and its low computational complexity. This work proposes an adaptive wavelet coding whose parameters of code rate and signal constellation can vary according to the fading level and evaluates the use of this transmission system in a control loop implemented by wireless sensor nodes. The performance of the adaptive system was evaluated in terms of bit error rate (BER versus E b / N 0 and spectral efficiency, considering a time-varying channel with flat Rayleigh fading, and in terms of processing overhead on a control system with wireless communication. The results obtained through computational simulations and experimental tests show performance gains obtained by insertion of the adaptive wavelet coding in a control loop with nodes interconnected by wireless link. These results enable the use of this technique in a wireless link control loop.

  6. Network coding and its applications to satellite systems

    DEFF Research Database (Denmark)

    Vieira, Fausto; Roetter, Daniel Enrique Lucani

    2015-01-01

    Network coding has its roots in information theory where it was initially proposed as a way to improve a two-node communication using a (broadcasting) relay. For this theoretical construct, a satellite communications system was proposed as an illustrative example, where the relay node would be a ...

  7. Consequences of the Solar System passage through dense interstellar clouds

    Directory of Open Access Journals (Sweden)

    A. G. Yeghikyan

    2003-06-01

    Full Text Available Several consequences of the passage of the solar system through dense interstellar molecular clouds are discussed. These clouds, dense (more than 100 cm-3, cold (10–50 K and extended (larger than 1 pc, are characterized by a gas-to-dust mass ratio of about 100, by a specific power grain size spectrum (grain radii usually cover the range 0.001–3 micron and by an average dust-to-gas number density ratio of about 10-12. Frequently these clouds contain small-scale (10–100 AU condensations with gas concentrations ranging up to 10 5 cm-3. At their casual passage over the solar system they exert pressures very much enhanced with respect to today’s standards. Under these conditions it will occur that the Earth is exposed directly to the interstellar flow. It is shown first that even close to the Sun, at 1 AU, the cloud’s matter is only partly ionized and should mainly interact with the solar wind by charge exchange processes. Dust particles of the cloud serve as a source of neutrals, generated by the solar UV irradiation of dust grains, causing the evaporation of icy materials. The release of neutral atoms from dust grains is then followed by strong influences on the solar wind plasma flow. The behavior of the neutral gas inflow parameters is investigated by a 2-D hydrodynamic approach to model the interaction processes. Because of a reduction of the heliospheric dimension down to 1 AU, direct influence of the cloud’s matter to the terrestrial environment and atmosphere could be envisaged.Key words. Interplanetary physics (heliopause and solar wind termination; interplanetary dust; interstellar gas

  8. Consequences of the Solar System passage through dense interstellar clouds

    Directory of Open Access Journals (Sweden)

    A. G. Yeghikyan

    Full Text Available Several consequences of the passage of the solar system through dense interstellar molecular clouds are discussed. These clouds, dense (more than 100 cm-3, cold (10–50 K and extended (larger than 1 pc, are characterized by a gas-to-dust mass ratio of about 100, by a specific power grain size spectrum (grain radii usually cover the range 0.001–3 micron and by an average dust-to-gas number density ratio of about 10-12. Frequently these clouds contain small-scale (10–100 AU condensations with gas concentrations ranging up to 10 5 cm-3. At their casual passage over the solar system they exert pressures very much enhanced with respect to today’s standards. Under these conditions it will occur that the Earth is exposed directly to the interstellar flow. It is shown first that even close to the Sun, at 1 AU, the cloud’s matter is only partly ionized and should mainly interact with the solar wind by charge exchange processes. Dust particles of the cloud serve as a source of neutrals, generated by the solar UV irradiation of dust grains, causing the evaporation of icy materials. The release of neutral atoms from dust grains is then followed by strong influences on the solar wind plasma flow. The behavior of the neutral gas inflow parameters is investigated by a 2-D hydrodynamic approach to model the interaction processes. Because of a reduction of the heliospheric dimension down to 1 AU, direct influence of the cloud’s matter to the terrestrial environment and atmosphere could be envisaged.

    Key words. Interplanetary physics (heliopause and solar wind termination; interplanetary dust; interstellar gas

  9. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  10. PCS a code system for generating production cross section libraries

    International Nuclear Information System (INIS)

    Cox, L.J.

    1997-01-01

    This document outlines the use of the PCS Code System. It summarizes the execution process for generating FORMAT2000 production cross section files from FORMAT2000 reaction cross section files. It also describes the process of assembling the ASCII versions of the high energy production files made from ENDL and Mark Chadwick's calculations. Descriptions of the function of each code along with its input and output and use are given. This document is under construction. Please submit entries, suggestions, questions, and corrections to (ljc at sign llnl.gov) 3 tabs

  11. PHITS-a particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji; Sato, Tatsuhiko; Iwase, Hiroshi; Nose, Hiroyuki; Nakashima, Hiroshi; Sihver, Lembit

    2006-01-01

    The paper presents a summary of the recent development of the multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS. In particular, we discuss in detail the development of two new models, JAM and JQMD, for high energy particle interactions, incorporated in PHITS, and show comparisons between model calculations and experiments for the validations of these models. The paper presents three applications of the code including spallation neutron source, heavy ion therapy and space radiation. The results and examples shown indicate PHITS has great ability of carrying out the radiation transport analysis of almost all particles including heavy ions within a wide energy range

  12. ERRORJ. Covariance processing code system for JENDL. Version 2

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-09-01

    ERRORJ is the covariance processing code system for Japanese Evaluated Nuclear Data Library (JENDL) that can produce group-averaged covariance data to apply it to the uncertainty analysis of nuclear characteristics. ERRORJ can treat the covariance data for cross sections including resonance parameters as well as angular distributions and energy distributions of secondary neutrons which could not be dealt with by former covariance processing codes. In addition, ERRORJ can treat various forms of multi-group cross section and produce multi-group covariance file with various formats. This document describes an outline of ERRORJ and how to use it. (author)

  13. Improved atmospheric dispersion modelling in the new program system UFOMOD for accident consequence assessments

    International Nuclear Information System (INIS)

    Panitz, H.J.

    1988-01-01

    An essential aim of the improvements of the new program system UFOMOD for Accident Consequence Assessments (ACAs) was to substitute the straightline Gaussian plume model conventionally used in ACA models by more realistic atmospheric dispersion models. To identify improved models which can be applied in ACA codes and to quantify the implications of different concepts of dispersion modelling on the results of an ACA, probabilistic comparative calculations with different atmospheric dispersion models have been carried out. The study showed that there are trajectory models available which can be applied in ACAs and that these trajectory models provide more realistic results of ACAs than straight-line Gaussian models. This led to a completly novel concept of atmospheric dispersion modelling which distinguish between two different distance ranges of validity: the near range ( 50 km). The two ranges are assigned to respective trajectory models

  14. Results from the Coded Aperture Neutron Imaging System (CANIS)

    International Nuclear Information System (INIS)

    Brubaker, Erik; Steele, John T.; Brennan, James S.; Hilton, Nathan R.; Marleau, Peter

    2010-01-01

    Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging- a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.

  15. Results from the coded aperture neutron imaging system

    International Nuclear Information System (INIS)

    Brubaker, Erik; Steele, John T.; Brennan, James S.; Marleau, Peter

    2010-01-01

    Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging - a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.

  16. Fusion PIC code performance analysis on the Cori KNL system

    Energy Technology Data Exchange (ETDEWEB)

    Koskela, Tuomas S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Deslippe, Jack [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Friesen, Brian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Raman, Karthic [INTEL Corp. (United States)

    2017-05-25

    We study the attainable performance of Particle-In-Cell codes on the Cori KNL system by analyzing a miniature particle push application based on the fusion PIC code XGC1. We start from the most basic building blocks of a PIC code and build up the complexity to identify the kernels that cost the most in performance and focus optimization efforts there. Particle push kernels operate at high AI and are not likely to be memory bandwidth or even cache bandwidth bound on KNL. Therefore, we see only minor benefits from the high bandwidth memory available on KNL, and achieving good vectorization is shown to be the most beneficial optimization path with theoretical yield of up to 8x speedup on KNL. In practice we are able to obtain up to a 4x gain from vectorization due to limitations set by the data layout and memory latency.

  17. Development of particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)

  18. The Vulnerability Assessment Code for Physical Protection System

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Yoo, Ho Sik

    2007-01-01

    To neutralize the increasing terror threats, nuclear facilities have strong physical protection system (PPS). PPS includes detectors, door locks, fences, regular guard patrols, and a hot line to a nearest military force. To design an efficient PPS and to fully operate it, vulnerability assessment process is required. Evaluating PPS of a nuclear facility is complicate process and, hence, several assessment codes have been developed. The estimation of adversary sequence interruption (EASI) code analyzes vulnerability along a single intrusion path. To evaluate many paths to a valuable asset in an actual facility, the systematic analysis of vulnerability to intrusion (SAVI) code was developed. KAERI improved SAVI and made the Korean analysis of vulnerability to intrusion (KAVI) code. Existing codes (SAVI and KAVI) have limitations in representing the distance of a facility because they use the simplified model of a PPS called adversary sequence diagram. In adversary sequence diagram the position of doors, sensors and fences is described just as the locating area. Thus, the distance between elements is inaccurate and we cannot reflect the range effect of sensors. In this abstract, we suggest accurate and intuitive vulnerability assessment based on raster map modeling of PPS. The raster map of PPS accurately represents the relative position of elements and, thus, the range effect of sensor can be easily incorporable. Most importantly, the raster map is easy to understand

  19. Scaling of Thermal-Hydraulic Phenomena and System Code Assessment

    International Nuclear Information System (INIS)

    Wolfert, K.

    2008-01-01

    In the last five decades large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Many separate effects tests and integral system tests were carried out to establish a data base for code development and code validation. In this context the question has to be answered, to what extent the results of down-scaled test facilities represent the thermal-hydraulic behaviour expected in a full-scale nuclear reactor under accidental conditions. Scaling principles, developed by many scientists and engineers, present a scientific technical basis and give a valuable orientation for the design of test facilities. However, it is impossible for a down-scaled facility to reproduce all physical phenomena in the correct temporal sequence and in the kind and strength of their occurrence. The designer needs to optimize a down-scaled facility for the processes of primary interest. This leads compulsorily to scaling distortions of other processes with less importance. Taking into account these weak points, a goal oriented code validation strategy is required, based on the analyses of separate effects tests and integral system tests as well as transients occurred in full-scale nuclear reactors. The CSNI validation matrices are an excellent basis for the fulfilling of this task. Separate effects tests in full scale play here an important role.

  20. Dose calculations for a simplified Mammosite system with the Monte Carlo Penelope and MCNPX simulation codes

    International Nuclear Information System (INIS)

    Rojas C, E.L.; Varon T, C.F.; Pedraza N, R.

    2007-01-01

    The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)

  1. Source Code Analysis Laboratory (SCALe) for Energy Delivery Systems

    Science.gov (United States)

    2010-12-01

    technical competence for the type of tests and calibrations SCALe undertakes. Testing and calibration laboratories that comply with ISO / IEC 17025 ...and exec t [ ISO / IEC 2005]. f a software system indicates that the SCALe analysis di by a CERT secure coding standard. Successful conforma antees that...to be more secure than non- systems. However, no study has yet been performed to p t ssment in accordance with ISO / IEC 17000: “a demonstr g to a

  2. Particle and heavy ion transport code system; PHITS

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

  3. Development of new two-dimensional spectral/spatial code based on dynamic cyclic shift code for OCDMA system

    Science.gov (United States)

    Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria

    2017-07-01

    In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bandini, G., E-mail: giacomino.bandini@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Polidori, M. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Gerschenfeld, A.; Pialla, D.; Li, S. [Commissariat à l’Energie Atomique (CEA) (France); Ma, W.M.; Kudinov, P.; Jeltsov, M.; Kööp, K. [Royal Institute of Technology (KTH) (Sweden); Huber, K.; Cheng, X.; Bruzzese, C.; Class, A.G.; Prill, D.P. [Karlsruhe Institute of Technology (KIT) (Germany); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Geffray, C.; Macian-Juan, R. [Technische Universität München (TUM) (Germany); Maas, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)

    2015-01-15

    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.

  5. LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I

    1985-07-01

    Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs.

  6. LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs

  7. Development of Integrated Code for Risk Assessment (INCORIA) for Physical Protection System

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Seo, Hyung Min; Yoo, Ho Sik

    2010-01-01

    A physical protection system (PPS) integrates people, procedures and equipment for the protection of assets or facilities against theft, sabotage or other malevolent human attacks. Among critical facilities, nuclear facilities and nuclear weapon sites require the highest level of PPS. After the September 11, 2001 terrorist attacks, international communities, including the IAEA, have made substantial efforts to protect nuclear material and nuclear facilities. The international flow on nuclear security is using the concept or risk assessment. The concept of risk assessment is firstly devised by nuclear safety people. They considered nuclear safety including its possible risk, which is the frequency of failure and possible consequence. Nuclear security people also considers security risk, which is the frequency of threat action, vulnerability, and consequences. The concept means that we should protect more when the credible threat exists and the possible radiological consequence is high. Even if there are several risk assessment methods of nuclear security, the application needs the help of tools because of a lot of calculation. It's also hard to find tools for whole phase of risk assessment. Several codes exist for the part of risk assessment. SAVI are used for vulnerability of PPS. Vital area identification code is used for consequence analysis. We are developing Integrated Code for Risk Assessment (INCORIA) to apply risk assessment methods for nuclear facilities. INCORIA evaluates PP-KINAC measures and generation tools for threat scenario. PP-KINAC is risk assessment measures for physical protection system developed by Hosik Yoo and is easy to apply. A threat scenario tool is used to generate threat scenario, which is used as one of input value to PP-KINAC measures

  8. An Expert System for the Development of Efficient Parallel Code

    Science.gov (United States)

    Jost, Gabriele; Chun, Robert; Jin, Hao-Qiang; Labarta, Jesus; Gimenez, Judit

    2004-01-01

    We have built the prototype of an expert system to assist the user in the development of efficient parallel code. The system was integrated into the parallel programming environment that is currently being developed at NASA Ames. The expert system interfaces to tools for automatic parallelization and performance analysis. It uses static program structure information and performance data in order to automatically determine causes of poor performance and to make suggestions for improvements. In this paper we give an overview of our programming environment, describe the prototype implementation of our expert system, and demonstrate its usefulness with several case studies.

  9. A guide to the AUS modular neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1987-04-01

    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

  10. Distributed magnetic field positioning system using code division multiple access

    Science.gov (United States)

    Prigge, Eric A. (Inventor)

    2003-01-01

    An apparatus and methods for a magnetic field positioning system use a fundamentally different, and advantageous, signal structure and multiple access method, known as Code Division Multiple Access (CDMA). This signal architecture, when combined with processing methods, leads to advantages over the existing technologies, especially when applied to a system with a large number of magnetic field generators (beacons). Beacons at known positions generate coded magnetic fields, and a magnetic sensor measures a sum field and decomposes it into component fields to determine the sensor position and orientation. The apparatus and methods can have a large `building-sized` coverage area. The system allows for numerous beacons to be distributed throughout an area at a number of different locations. A method to estimate position and attitude, with no prior knowledge, uses dipole fields produced by these beacons in different locations.

  11. Improving system modeling accuracy with Monte Carlo codes

    International Nuclear Information System (INIS)

    Johnson, A.S.

    1996-01-01

    The use of computer codes based on Monte Carlo methods to perform criticality calculations has become common-place. Although results frequently published in the literature report calculated k eff values to four decimal places, people who use the codes in their everyday work say that they only believe the first two decimal places of any result. The lack of confidence in the computed k eff values may be due to the tendency of the reported standard deviation to underestimate errors associated with the Monte Carlo process. The standard deviation as reported by the codes is the standard deviation of the mean of the k eff values for individual generations in the computer simulation, not the standard deviation of the computed k eff value compared with the physical system. A more subtle problem with the standard deviation of the mean as reported by the codes is that all the k eff values from the separate generations are not statistically independent since the k eff of a given generation is a function of k eff of the previous generation, which is ultimately based on the starting source. To produce a standard deviation that is more representative of the physical system, statistically independent values of k eff are needed

  12. Coded aperture material motion detection system for the ACPR

    International Nuclear Information System (INIS)

    McArthur, D.A.; Kelly, J.G.

    1975-01-01

    Single LMFBR fuel pins are being irradiated in Sandia's Annular Core Pulsed Reactor (ACPR). In these experiments single fuel pins have been driven well into the melt and vaporization regions in transients with pulse widths of about 5 ms. The ACPR is being upgraded so that it can be used to irradiate bundles of seven LMFBR fuel pins. The coded aperture material motion detection system described is being developed for this upgraded ACPR, and has for its design goals 1 mm transverse resolution (i.e., in the axial and radial directions), depth resolution of a few cm, and time resolution of 0.1 ms. The target date for development of this system is fall 1977. The paper briefly reviews the properties of coded aperture imaging, describes one possible system for the ACPR upgrade, discusses experiments which have been performed to investigate the feasibility of such a system, and describes briefly the further work required to develop such a system. The type of coded aperture to be used has not yet been fixed, but a one-dimensional section of a Fresnel zone plate appears at this time to have significant advantages

  13. Development of computing code system for level 3 PSA

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan.

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  14. Development of computing code system for level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  15. Study on a new meteorological sampling scheme developed for the OSCAAR code system

    International Nuclear Information System (INIS)

    Liu Xinhe; Tomita, Kenichi; Homma, Toshimitsu

    2002-03-01

    One important step in Level-3 Probabilistic Safety Assessment is meteorological sequence sampling, on which the previous studies were mainly related to code systems using the straight-line plume model and more efforts are needed for those using the trajectory puff model such as the OSCAAR code system. This report describes the development of a new meteorological sampling scheme for the OSCAAR code system that explicitly considers population distribution. A group of principles set for the development of this new sampling scheme includes completeness, appropriate stratification, optimum allocation, practicability and so on. In this report, discussions are made about the procedures of the new sampling scheme and its application. The calculation results illustrate that although it is quite difficult to optimize stratification of meteorological sequences based on a few environmental parameters the new scheme do gather the most inverse conditions in a single subset of meteorological sequences. The size of this subset may be as small as a few dozens, so that the tail of a complementary cumulative distribution function is possible to remain relatively static in different trials of the probabilistic consequence assessment code. (author)

  16. A new two dimensional spectral/spatial multi-diagonal code for noncoherent optical code division multiple access (OCDMA) systems

    Science.gov (United States)

    Kadhim, Rasim Azeez; Fadhil, Hilal Adnan; Aljunid, S. A.; Razalli, Mohamad Shahrazel

    2014-10-01

    A new two dimensional codes family, namely two dimensional multi-diagonal (2D-MD) codes, is proposed for spectral/spatial non-coherent OCDMA systems based on the one dimensional MD code. Since the MD code has the property of zero cross correlation, the proposed 2D-MD code also has this property. So that, the multi-access interference (MAI) is fully eliminated and the phase induced intensity noise (PIIN) is suppressed with the proposed code. Code performance is analyzed in terms of bit error rate (BER) while considering the effect of shot noise, PIIN, and thermal noise. The performance of the proposed code is compared with the related MD, modified quadratic congruence (MQC), two dimensional perfect difference (2D-PD) and two dimensional diluted perfect difference (2D-DPD) codes. The analytical and the simulation results reveal that the proposed 2D-MD code outperforms the other codes. Moreover, a large number of simultaneous users can be accommodated at low BER and high data rate.

  17. Computer codes and methods for simulating accelerator driven systems

    International Nuclear Information System (INIS)

    Sartori, E.; Byung Chan Na

    2003-01-01

    A large set of computer codes and associated data libraries have been developed by nuclear research and industry over the past half century. A large number of them are in the public domain and can be obtained under agreed conditions from different Information Centres. The areas covered comprise: basic nuclear data and models, reactor spectra and cell calculations, static and dynamic reactor analysis, criticality, radiation shielding, dosimetry and material damage, fuel behaviour, safety and hazard analysis, heat conduction and fluid flow in reactor systems, spent fuel and waste management (handling, transportation, and storage), economics of fuel cycles, impact on the environment of nuclear activities etc. These codes and models have been developed mostly for critical systems used for research or power generation and other technological applications. Many of them have not been designed for accelerator driven systems (ADS), but with competent use, they can be used for studying such systems or can form the basis for adapting existing methods to the specific needs of ADS's. The present paper describes the types of methods, codes and associated data available and their role in the applications. It provides Web addresses for facilitating searches for such tools. Some indications are given on the effect of non appropriate or 'blind' use of existing tools to ADS. Reference is made to available experimental data that can be used for validating the methods use. Finally, some international activities linked to the different computational aspects are described briefly. (author)

  18. 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

  19. 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

  20. 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.

  1. Photovoltaic Power Systems and the National Electrical Code: Suggested Practices

    Energy Technology Data Exchange (ETDEWEB)

    None

    2002-02-01

    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.

  2. 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...

  3. Structure and operation of the ITS code system

    International Nuclear Information System (INIS)

    Halbleib, J.

    1988-01-01

    The TIGER series of time-independent coupled electron-photon Monte Carlo transport codes is a group of multimaterial and multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron-photon cascade by combining microscopic photon transport with a macroscopic random walk for electron transport. Major contributors to its evolution are listed. The author and his associates are primarily code users rather than code developers, and have borrowed freely from existing work wherever possible. Nevertheless, their efforts have resulted in various software packages for describing the production and transport of the electron-photon cascade that they found sufficiently useful to warrant dissemination through the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory. The ITS system (Integrated TIGER Series) represents the organization and integration of this combined software, along with much additional capability from previously unreleased work, into a single convenient package of exceptional user friendliness and portability. Emphasis is on simplicity and flexibility of application without sacrificing the rigor or sophistication of the physical model

  4. A seismic data compression system using subband coding

    Science.gov (United States)

    Kiely, A. B.; Pollara, F.

    1995-01-01

    This article presents a study of seismic data compression techniques and a compression algorithm based on subband coding. The algorithm includes three stages: a decorrelation stage, a quantization stage that introduces a controlled amount of distortion to allow for high compression ratios, and a lossless entropy coding stage based on a simple but efficient arithmetic coding method. Subband coding methods are particularly suited to the decorrelation of nonstationary processes such as seismic events. Adaptivity to the nonstationary behavior of the waveform is achieved by dividing the data into separate blocks that are encoded separately with an adaptive arithmetic encoder. This is done with high efficiency due to the low overhead introduced by the arithmetic encoder in specifying its parameters. The technique could be used as a progressive transmission system, where successive refinements of the data can be requested by the user. This allows seismologists to first examine a coarse version of waveforms with minimal usage of the channel and then decide where refinements are required. Rate-distortion performance results are presented and comparisons are made with two block transform methods.

  5. SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi

    2009-05-01

    Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)

  6. Multiple Description Coding for Closed Loop Systems over Erasure Channels

    DEFF Research Database (Denmark)

    Østergaard, Jan; Quevedo, Daniel

    2013-01-01

    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......) 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...... by showing that the system can be cast as a Markov jump linear system....

  7. Modeling of pipe break accident in a district heating system using RELAP5 computer code

    International Nuclear Information System (INIS)

    Kaliatka, A.; Valinčius, M.

    2012-01-01

    Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

  8. BER performance comparison of optical CDMA systems with/without turbo codes

    Science.gov (United States)

    Kulkarni, Muralidhar; Chauhan, Vijender S.; Dutta, Yashpal; Sinha, Ravindra K.

    2002-08-01

    In this paper, we have analyzed and simulated the BER performance of a turbo coded optical code-division multiple-access (TC-OCDMA) system. A performance comparison has been made between uncoded OCDMA and TC-OCDMA systems employing various OCDMA address codes (optical orthogonal codes (OOCs), Generalized Multiwavelength Prime codes (GMWPC's), and Generalized Multiwavelength Reed Solomon code (GMWRSC's)). The BER performance of TC-OCDMA systems has been analyzed and simulated by varying the code weight of address code employed by the system. From the simulation results, it is observed that lower weight address codes can be employed for TC-OCDMA systems that can have the equivalent BER performance of uncoded systems employing higher weight address codes for a fixed number of active users.

  9. Method of laser beam coding for control systems

    Science.gov (United States)

    Pałys, Tomasz; Arciuch, Artur; Walczak, Andrzej; Murawski, Krzysztof

    2017-08-01

    The article presents the method of encoding a laser beam for control systems. The experiments were performed using a red laser emitting source with a wavelength of λ = 650 nm and a power of P ≍ 3 mW. The aim of the study was to develop methods of modulation and demodulation of the laser beam. Results of research, in which we determined the effect of selected camera parameters, such as image resolution, number of frames per second on the result of demodulation of optical signal, is also shown in the paper. The experiments showed that the adopted coding method provides sufficient information encoded in a single laser beam (36 codes with the effectiveness of decoding at 99.9%).

  10. 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

    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 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....... 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....

  11. VACOSS - variable coding seal system for nuclear material control

    International Nuclear Information System (INIS)

    Kennepohl, K.; Stein, G.

    1977-12-01

    VACOSS - Variable Coding Seal System - is intended to seal: rooms and containers with nuclear material, nuclear instrumentation and equipment of the operator, instrumentation and equipment at the supervisory authority. It is easy to handle, reusable, transportable and consists of three components: 1. Seal. The light guide in fibre optics with infrared light emitter and receiver serves as lead. The statistical treatment of coded data given in the seal via adapter box guarantees an extremely high degree of access reliability. It is possible to store the data of two undue seal openings together with data concerning time and duration of the opening. 2. The adapter box can be used for input or input and output of data indicating the seal integrity. 3. The simulation programme is located in the computing center of the supervisory authority and permits to determine date and time of opening by decoding the seal memory data. (orig./WB) [de

  12. Alignment effects on a neutron imaging system using coded apertures

    International Nuclear Information System (INIS)

    Thfoin, Isabelle; Landoas, Olivier; Caillaud, Tony; Vincent, Maxime; Bourgade, Jean-Luc; Rosse, Bertrand; Disdier, Laurent; Sangster, Thomas C.; Glebov, Vladimir Yu.; Pien, Greg; Armstrong, William

    2010-01-01

    A high resolution neutron imaging system is being developed and tested on the OMEGA laser facility for inertial confinement fusion experiments. This diagnostic uses a coded imaging technique with a penumbral or an annular aperture. The sensitiveness of these techniques to misalignment was pointed out with both experiments and simulations. Results obtained during OMEGA shots are in good agreement with calculations performed with the Monte Carlo code GEANT4. Both techniques are sensitive to the relative position of the source in the field of view. The penumbral imaging technique then demonstrates to be less sensitive to misalignment compared to the ring. These results show the necessity to develop a neutron imaging diagnostic for megajoule class lasers taking into account our alignment capabilities on such facilities.

  13. 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.

  14. Transmission over UWB channels with OFDM system using LDPC coding

    Science.gov (United States)

    Dziwoki, Grzegorz; Kucharczyk, Marcin; Sulek, Wojciech

    2009-06-01

    Hostile wireless environment requires use of sophisticated signal processing methods. The paper concerns on Ultra Wideband (UWB) transmission over Personal Area Networks (PAN) including MB-OFDM specification of physical layer. In presented work the transmission system with OFDM modulation was connected with LDPC encoder/decoder. Additionally the frame and bit error rate (FER and BER) of the system was decreased using results from the LDPC decoder in a kind of turbo equalization algorithm for better channel estimation. Computational block using evolutionary strategy, from genetic algorithms family, was also used in presented system. It was placed after SPA (Sum-Product Algorithm) decoder and is conditionally turned on in the decoding process. The result is increased effectiveness of the whole system, especially lower FER. The system was tested with two types of LDPC codes, depending on type of parity check matrices: randomly generated and constructed deterministically, optimized for practical decoder architecture implemented in the FPGA device.

  15. Development of a domestically-made system code

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from the Fukushima-Daiichi NPP accidents, a new safety standard based on state-of-the-art findings has been established by the Japanese Nuclear Regulation Authority (NRA) and will soon come into force in Japan. In order to ensure a precise response to this movement from a technological point of view, it should be required for safety regulation to develop a new system code with much smaller uncertainty and reinforced simulation capability even in application to beyond-DBAs (BDBAs), as well as with the capability of close coupling to a newly developing severe accident code. Accordingly, development of a new domestically-made system code that incorporates 3-dimensional and 3 or more fluid thermal-hydraulics in tandem with a 3-dimensional neutronics has been started in 2012. In 2012, two branches of development activities, the development of 'main body' and advanced features have been started in parallel for development efficiency. The main body has been started from scratch and the following activities have therefore been performed: 1) development and determination of key principles and methodologies to realize a flexible, extensible and robust platform, 2) determination of requirements definition, 3) start of basic program design and coding and 4) start of a development of prototypical GUI-based pre-post processor. As for the advanced features, the following activities have been performed: 1) development of Phenomena Identification and Ranking Tables (PIRTs) and model capability matrix from normal operations to BDBAs in order to address requirements definition for advanced modeling, 2) development of detailed action plan for modification of field equations, numerical schemes and solvers and 3) start of the program development of field equations with an interfacial area concentration transport equation, a robust solver for condensation induced water hammer phenomena and a versatile Newton-Raphson solver. (author)

  16. A guide to TIRION 4 - a computer code for calculating the consequences of releasing radioactive material to the atmosphere

    International Nuclear Information System (INIS)

    Fryer, L.S.

    1978-12-01

    TIRION 4 is the most recent program in a series designed to calculate the consequences of releasing radioactive material to the atmosphere. A brief description of the models used in the program and full details of the various control cards necessary to run TIRION 4 are given. (author)

  17. Performance enhancement of successive interference cancellation scheme based on spectral amplitude coding for optical code-division multiple-access systems using Hadamard codes

    Science.gov (United States)

    Eltaif, Tawfig; Shalaby, Hossam M. H.; Shaari, Sahbudin; Hamarsheh, Mohammad M. N.

    2009-04-01

    A successive interference cancellation scheme is applied to optical code-division multiple-access (OCDMA) systems with spectral amplitude coding (SAC). A detailed analysis of this system, with Hadamard codes used as signature sequences, is presented. The system can easily remove the effect of the strongest signal at each stage of the cancellation process. In addition, simulation of the prose system is performed in order to validate the theoretical results. The system shows a small bit error rate at a large number of active users compared to the SAC OCDMA system. Our results reveal that the proposed system is efficient in eliminating the effect of the multiple-user interference and in the enhancement of the overall performance.

  18. A study on the nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Lee, Jong Bok; Choi, Young Gil; Suh, Soong Hyok; Kang, Byong Heon; Kim, Hee Kyung; Kim, Ko Ryeo; Park, Soo Jin

    1990-12-01

    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)

  19. Athermalization of infrared dual field optical system based on wavefront coding

    Science.gov (United States)

    Jiang, Kai; Jiang, Bo; Liu, Kai; Yan, Peipei; Duan, Jing; Shan, Qiu-sha

    2017-02-01

    Wavefront coding is a technology which combination of the optical design and digital image processing. By inserting a phase mask closed to the pupil plane of the optical system the wavefront of the system is re-modulated. And the depth of focus is extended consequently. In reality the idea is same as the athermalization theory of infrared optical system. In this paper, an uncooled infrared dual field optical system with effective focal as 38mm/19mm, F number as 1.2 of both focal length, operating wavelength varying from 8μm to 12μm was designed. A cubic phase mask was used at the pupil plane to re-modulate the wavefront. Then the performance of the infrared system was simulated with CODEV as the environment temperature varying from -40° to 60°. MTF curve of the optical system with phase mask are compared with the outcome before using phase mask. The result show that wavefront coding technology can make the system not sensitive to thermal defocus, and then realize the athermal design of the infrared optical system.

  20. Performance and Complexity Evaluation of Iterative Receiver for Coded MIMO-OFDM Systems

    Directory of Open Access Journals (Sweden)

    Rida El Chall

    2016-01-01

    Full Text Available Multiple-input multiple-output (MIMO technology in combination with channel coding technique is a promising solution for reliable high data rate transmission in future wireless communication systems. However, these technologies pose significant challenges for the design of an iterative receiver. In this paper, an efficient receiver combining soft-input soft-output (SISO detection based on low-complexity K-Best (LC-K-Best decoder with various forward error correction codes, namely, LTE turbo decoder and LDPC decoder, is investigated. We first investigate the convergence behaviors of the iterative MIMO receivers to determine the required inner and outer iterations. Consequently, the performance of LC-K-Best based receiver is evaluated in various LTE channel environments and compared with other MIMO detection schemes. Moreover, the computational complexity of the iterative receiver with different channel coding techniques is evaluated and compared with different modulation orders and coding rates. Simulation results show that LC-K-Best based receiver achieves satisfactory performance-complexity trade-offs.

  1. System code improvements for modelling passive safety systems and their validation

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

  2. An engineering code to analyze hypersonic thermal management systems

    Science.gov (United States)

    Vangriethuysen, Valerie J.; Wallace, Clark E.

    1993-01-01

    Thermal loads on current and future aircraft are increasing and as a result are stressing the energy collection, control, and dissipation capabilities of current thermal management systems and technology. The thermal loads for hypersonic vehicles will be no exception. In fact, with their projected high heat loads and fluxes, hypersonic vehicles are a prime example of systems that will require thermal management systems (TMS) that have been optimized and integrated with the entire vehicle to the maximum extent possible during the initial design stages. This will not only be to meet operational requirements, but also to fulfill weight and performance constraints in order for the vehicle to takeoff and complete its mission successfully. To meet this challenge, the TMS can no longer be two or more entirely independent systems, nor can thermal management be an after thought in the design process, the typical pervasive approach in the past. Instead, a TMS that was integrated throughout the entire vehicle and subsequently optimized will be required. To accomplish this, a method that iteratively optimizes the TMS throughout the vehicle will not only be highly desirable, but advantageous in order to reduce the manhours normally required to conduct the necessary tradeoff studies and comparisons. A thermal management engineering computer code that is under development and being managed at Wright Laboratory, Wright-Patterson AFB, is discussed. The primary goal of the code is to aid in the development of a hypersonic vehicle TMS that has been optimized and integrated on a total vehicle basis.

  3. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2002-03-01

    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)

  4. Simplified modeling and code usage in the PASC-3 code system by the introduction of a programming environment

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Oppe, J.; Oudshoorn, H.L.; Slobben, J.

    1991-06-01

    A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified. Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab

  5. SALT [System Analysis Language Translater]: A steady state and dynamic systems code

    International Nuclear Information System (INIS)

    Berry, G.; Geyer, H.

    1983-01-01

    SALT (System Analysis Language Translater) is a lumped parameter approach to system analysis which is totally modular. The modules are all precompiled and only the main program, which is generated by SALT, needs to be compiled for each unique system configuration. This is a departure from other lumped parameter codes where all models are written by MACROS and then compiled for each unique configuration, usually after all of the models are lumped together and sorted to eliminate undetermined variables. The SALT code contains a robust and sophisticated steady-sate finder (non-linear equation solver), optimization capability and enhanced GEAR integration scheme which makes use of sparsity and algebraic constraints. The SALT systems code has been used for various technologies. The code was originally developed for open-cycle magnetohydrodynamic (MHD) systems. It was easily extended to liquid metal MHD systems by simply adding the appropriate models and property libraries. Similarly, the model and property libraries were expanded to handle fuel cell systems, flue gas desulfurization systems, combined cycle gasification systems, fluidized bed combustion systems, ocean thermal energy conversion systems, geothermal systems, nuclear systems, and conventional coal-fired power plants. Obviously, the SALT systems code is extremely flexible to be able to handle all of these diverse systems. At present, the dynamic option has only been used for LMFBR nuclear power plants and geothermal power plants. However, it can easily be extended to other systems and can be used for analyzing control problems. 12 refs

  6. Assessment of Radiological and Economic Consequences of a Hypothetical Accident for ETRR-2, Egypt Utilizing COSYMA Code

    International Nuclear Information System (INIS)

    Tawfik, F.S.; Abdel-Aal, M.M.

    2008-01-01

    A comprehensive probabilistic study of an accident consequence assessment (ACA) for loss of coolant accident (LOCA) has accomplished to the second research reactor ETRR-2, located at Inshas Nuclear Research Center, Cairo, Egypt. PC-COSYMA, developed with the support of European Commission, has adopted to assess the radiological and economic consequences of a proposed accident. The consequences of the accident evaluated in case of early and late effects. The effective doses and doses in different organs carried out with and without countermeasures. The force mentioned calculations were required the following studies: the core inventory due to the hypothetical accident, the physical parameters of the source term, the hourly basis meteorological parameters for one complete year, and the population distribution around the plant. The hourly stability conditions and height of atmospheric boundary layers (ABL) of the concerned site were calculated. The results showed that, the nuclides that have short half-lives (few days) give the highest air and ground concentrations after the accident than the others. The area around the reactor requires the early and late countermeasures action after the accident especially in the downwind sectors. Economically, the costs of emergency plan are effectively high in case of applying countermeasures but countermeasures reduce the risk effects

  7. 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

  8. PWR core follow calculations using the ELCOS code system

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1990-01-01

    The ELCOS code system developed at PSI is used to simulate a cycle of a PWR in which one fifth of the assemblies are MOX fuel. The reactor and the calculational methods are briefly described. The calculated critical boron concentrations and power distributions are compared with the measurements at the plant. Although the critical boron concentration is somewhat overpredicted and the computed power distributions are slightly flatter than the measured ones the results of the calculations agree generally well with the measured data. (author) 1 tab., 8 figs., 6 refs

  9. 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...

  10. Geographic Information Systems using CODES linked data (Crash outcome data evaluation system)

    Science.gov (United States)

    2001-04-01

    This report presents information about geographic information systems (GIS) and CODES linked data. Section one provides an overview of a GIS and the benefits of linking to CODES. Section two outlines the basic issues relative to the types of map data...

  11. [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

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

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

    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

  13. Evaluation of system codes for analyzing naturally circulating gas loop

    International Nuclear Information System (INIS)

    Lee, Jeong Ik; No, Hee Cheon; Hejzlar, Pavel

    2009-01-01

    Steady-state natural circulation data obtained in a 7 m-tall experimental loop with carbon dioxide and nitrogen are presented in this paper. The loop was originally designed to encompass operating range of a prototype gas-cooled fast reactor passive decay heat removal system, but the results and conclusions are applicable to any natural circulation loop operating in regimes having buoyancy and acceleration parameters within the ranges validated in this loop. Natural circulation steady-state data are compared to numerical predictions by two system analysis codes: GAMMA and RELAP5-3D. GAMMA is a computational tool for predicting various transients which can potentially occur in a gas-cooled reactor. The code has a capability of analyzing multi-dimensional multi-component mixtures and includes models for friction, heat transfer, chemical reaction, and multi-component molecular diffusion. Natural circulation data with two gases show that the loop operates in the deteriorated turbulent heat transfer (DTHT) regime which exhibits substantially reduced heat transfer coefficients compared to the forced turbulent flow. The GAMMA code with an original heat transfer package predicted conservative results in terms of peak wall temperature. However, the estimated peak location did not successfully match the data. Even though GAMMA's original heat transfer package included mixed-convection regime, which is a part of the DTHT regime, the results showed that the original heat transfer package could not reproduce the data with sufficient accuracy. After implementing a recently developed correlation and corresponding heat transfer regime map into GAMMA to cover the whole range of the DTHT regime, we obtained better agreement with the data. RELAP5-3D results are discussed in parallel.

  14. Coherence of Radial Implicative Fuzzy Systems with Nominal Consequents

    Czech Academy of Sciences Publication Activity Database

    Coufal, David

    -, č. 4 (2006), s. 60-66 ISSN 1509-4553 R&D Projects: GA MŠk 1M0545 Institutional research plan: CEZ:AV0Z10300504 Keywords : implicative fuzzy system * radial fuzzy system * nominal output space * coherence Subject RIV: IN - Informatics, Computer Science

  15. Economic consequences of investing in sensor systems on dairy farms

    NARCIS (Netherlands)

    Steeneveld, W.; Hogeveen, H.; Oude Lansink, A.G.J.M.

    2015-01-01

    The objective of this study was to investigate the impact of investment in sensor systems on productivity change, using farm accounting data. Farm accounting data for the years 2008–2013 was available for 217 Dutch dairy farms. In addition, information was available on the adoption of sensor systems

  16. Abstract feature codes: The building blocks of the implicit learning system.

    Science.gov (United States)

    Eberhardt, Katharina; Esser, Sarah; Haider, Hilde

    2017-07-01

    According to the Theory of Event Coding (TEC; Hommel, Müsseler, Aschersleben, & Prinz, 2001), action and perception are represented in a shared format in the cognitive system by means of feature codes. In implicit sequence learning research, it is still common to make a conceptual difference between independent motor and perceptual sequences. This supposedly independent learning takes place in encapsulated modules (Keele, Ivry, Mayr, Hazeltine, & Heuer 2003) that process information along single dimensions. These dimensions have remained underspecified so far. It is especially not clear whether stimulus and response characteristics are processed in separate modules. Here, we suggest that feature dimensions as they are described in the TEC should be viewed as the basic content of modules of implicit learning. This means that the modules process all stimulus and response information related to certain feature dimensions of the perceptual environment. In 3 experiments, we investigated by means of a serial reaction time task the nature of the basic units of implicit learning. As a test case, we used stimulus location sequence learning. The results show that a stimulus location sequence and a response location sequence cannot be learned without interference (Experiment 2) unless one of the sequences can be coded via an alternative, nonspatial dimension (Experiment 3). These results support the notion that spatial location is one module of the implicit learning system and, consequently, that there are no separate processing units for stimulus versus response locations. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  17. Development of GUI systems for the MIDAS code

    International Nuclear Information System (INIS)

    Kim, K.R.; Park, S.H.; Kim, D.H.

    2004-01-01

    MIDAS is being developed at KAERI based on MELCOR as an integrated severe accident analysis code with existing model modification and new model addition. MIDAS was restructured to avoid the pointer based variable referencing style of MELCOR, and enhanced the memory effectiveness using the dynamic allocation method of Fortran 90. This paper describes recent activities of developing the GUI environments for MIDAS code at KAERI. Up to now, we have developed the four PC-based subsystems, which are IEDIT, IPLOT, SATS and HyperKAMG. IEDIT is an input management system that can read MELCOR input files and display its information in the Window panels. Users can modify each item in the panel and the input file will be modified according to that changes. IPLOT is a simple plotting system that can draw MIDAS plot variables trend graphs. SATS is developed as a severe accident training simulator that can display nuclear plant behavior graphically. Moreover SATS provides several controllable pumps and valves which appeared in the severe accidence. Together with SATS and the online severe accident guidance HyperKAMG, combined properly, severe accident mitigation scenarios could be presented graphically and dramatically without any change of MELCOR inputs. GUI development as a part of a severe accident management program package, MIDAS. (author)

  18. Source Code Verification for Embedded Systems using Prolog

    Directory of Open Access Journals (Sweden)

    Frank Flederer

    2017-01-01

    Full Text Available System relevant embedded software needs to be reliable and, therefore, well tested, especially for aerospace systems. A common technique to verify programs is the analysis of their abstract syntax tree (AST. Tree structures can be elegantly analyzed with the logic programming language Prolog. Moreover, Prolog offers further advantages for a thorough analysis: On the one hand, it natively provides versatile options to efficiently process tree or graph data structures. On the other hand, Prolog's non-determinism and backtracking eases tests of different variations of the program flow without big effort. A rule-based approach with Prolog allows to characterize the verification goals in a concise and declarative way. In this paper, we describe our approach to verify the source code of a flash file system with the help of Prolog. The flash file system is written in C++ and has been developed particularly for the use in satellites. We transform a given abstract syntax tree of C++ source code into Prolog facts and derive the call graph and the execution sequence (tree, which then are further tested against verification goals. The different program flow branching due to control structures is derived by backtracking as subtrees of the full execution sequence. Finally, these subtrees are verified in Prolog. We illustrate our approach with a case study, where we search for incorrect applications of semaphores in embedded software using the real-time operating system RODOS. We rely on computation tree logic (CTL and have designed an embedded domain specific language (DSL in Prolog to express the verification goals.

  19. Simulation for photon detection in spectrometric system of high purity (HPGe) using MCNPX code

    International Nuclear Information System (INIS)

    Correa, Guilherme Jorge de Souza

    2013-01-01

    The Brazilian National Commission of Nuclear Energy defines parameters for classification and management of radioactive waste in accordance with the activity of materials. The efficiency of a detection system is crucial to determine the real activity of a radioactive source. When it's possible, the system's calibration should be performed using a standard source. Unfortunately, there are only a few cases that it can be done this way, considering the difficulty of obtaining appropriate standard sources for each type of measurement. So, computer simulations can be performed to assist in calculating of the efficiency of the system and, consequently, also auxiliary the classification of radioactive waste. This study aims to model a high purity germanium (HPGe) detector with MCNPX code, approaching the spectral values computationally obtained of the values experimentally obtained for the photopeak of 137 Cs. The approach will be made through changes in outer dead layer of the germanium crystal modeled. (author)

  20. Interval Coded Scoring: a toolbox for interpretable scoring systems

    Directory of Open Access Journals (Sweden)

    Lieven Billiet

    2018-04-01

    Full Text Available Over the last decades, clinical decision support systems have been gaining importance. They help clinicians to make effective use of the overload of available information to obtain correct diagnoses and appropriate treatments. However, their power often comes at the cost of a black box model which cannot be interpreted easily. This interpretability is of paramount importance in a medical setting with regard to trust and (legal responsibility. In contrast, existing medical scoring systems are easy to understand and use, but they are often a simplified rule-of-thumb summary of previous medical experience rather than a well-founded system based on available data. Interval Coded Scoring (ICS connects these two approaches, exploiting the power of sparse optimization to derive scoring systems from training data. The presented toolbox interface makes this theory easily applicable to both small and large datasets. It contains two possible problem formulations based on linear programming or elastic net. Both allow to construct a model for a binary classification problem and establish risk profiles that can be used for future diagnosis. All of this requires only a few lines of code. ICS differs from standard machine learning through its model consisting of interpretable main effects and interactions. Furthermore, insertion of expert knowledge is possible because the training can be semi-automatic. This allows end users to make a trade-off between complexity and performance based on cross-validation results and expert knowledge. Additionally, the toolbox offers an accessible way to assess classification performance via accuracy and the ROC curve, whereas the calibration of the risk profile can be evaluated via a calibration curve. Finally, the colour-coded model visualization has particular appeal if one wants to apply ICS manually on new observations, as well as for validation by experts in the specific application domains. The validity and applicability

  1. 42 CFR 405.512 - Carriers' procedural terminology and coding systems.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 2 2010-10-01 2010-10-01 false Carriers' procedural terminology and coding systems... Determining Reasonable Charges § 405.512 Carriers' procedural terminology and coding systems. (a) General. Procedural terminology and coding systems are designed to provide physicians and third party payers with a...

  2. Development of EASYQAD version β: A Visualization Code System for QAD-CGGP-A Gamma and Neutron Shielding Calculation Code

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung

    2007-01-01

    EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system

  3. Aqueous Transport Code Revisions Using Geographic Information Systems

    International Nuclear Information System (INIS)

    Chen, K.F.

    2003-01-01

    STREAM II, developed at the Savannah River Site (SRS) for execution on a personal computer, is an emergency response code that predicts downstream pollutant concentrations for releases from the SRS area to the Savannah River for emergency response management decision making. The STREAM II code consists of pre-processor, calculation, and post-processor modules. The pre-processor module provides a graphical user interface (GUI) for inputting the initial release data. The GUI passes the user specified data to the calculation module that calculates the pollutant concentrations at downstream locations and the transport times. The calculation module of the STREAM II adopts the transport module of the WASP5 code. WASP5 is a US Environmental Protection Agency water quality analysis program that simulates pollutant transport and fate through surface water using a finite difference method to solve the transport equation. The calculated downstream pollutant concentrations and travel times a re passed to the post-processor for display on the computer screen in graphical and tabular forms. To minimize the user's effort in the emergency situation, the required input parameters are limited to the time and date of release, type of release, location of release, amount and duration of release, and the calculation units. The user, however, could only select one of the seventeen predetermined locations. Hence, STREAM II could not be used for situations in which release locations differ from the seventeen predetermined locations. To eliminate this limitation, STREAM II has been revised to allow users to select the release location anywhere along the specified SRS main streams or the Savannah River by mouse-selection from a map displayed on the computer monitor. The required modifications to STREAM II using geographic information systems (GIS) software is discussed in this paper

  4. Product Configuration Systems and Consequences for Productivity and Quality

    DEFF Research Database (Denmark)

    Pedersen, Jørgen Lindgaard; Edwards, Kasper

    2004-01-01

    with informations about expected and realised costs and benefits in implemented PCS's we have the following results: the three highest aggregated scoring expected benefits are: 1)improved quality in specifications, 2)lower turnaround time, and 3)using less resources.The results achieved are in accordance......Product Configuration Systems (PCS) are systems to produce quotations to potential buyers of the products an enterprise produces. Typically the products are complex and will first be produced according to specific needs from the customers. From twelve case studies in Danish industry...

  5. The mirror neuron system and the consequences of its dysfunction.

    Science.gov (United States)

    Iacoboni, Marco; Dapretto, Mirella

    2006-12-01

    The discovery of premotor and parietal cells known as mirror neurons in the macaque brain that fire not only when the animal is in action, but also when it observes others carrying out the same actions provides a plausible neurophysiological mechanism for a variety of important social behaviours, from imitation to empathy. Recent data also show that dysfunction of the mirror neuron system in humans might be a core deficit in autism, a socially isolating condition. Here, we review the neurophysiology of the mirror neuron system and its role in social cognition and discuss the clinical implications of mirror neuron dysfunction.

  6. Design of ACM system based on non-greedy punctured LDPC codes

    Science.gov (United States)

    Lu, Zijun; Jiang, Zihong; Zhou, Lin; He, Yucheng

    2017-08-01

    In this paper, an adaptive coded modulation (ACM) scheme based on rate-compatible LDPC (RC-LDPC) codes was designed. The RC-LDPC codes were constructed by a non-greedy puncturing method which showed good performance in high code rate region. Moreover, the incremental redundancy scheme of LDPC-based ACM system over AWGN channel was proposed. By this scheme, code rates vary from 2/3 to 5/6 and the complication of the ACM system is lowered. Simulations show that more and more obvious coding gain can be obtained by the proposed ACM system with higher throughput.

  7. [Aging of the respiratory system: anatomical changes and physiological consequences].

    Science.gov (United States)

    Ketata, W; Rekik, W K; Ayadi, H; Kammoun, S

    2012-10-01

    The respiratory system undergoes progressive involution with age, resulting in anatomical and functional changes that are exerted on all levels. The rib cage stiffens and respiratory muscles weaken. Distal bronchioles have reduced diameter and tend to be collapsed. Mobilized lung volumes decrease with age while residual volume increases. Gas exchanges are modified with a linear decrease of PaO(2) up to the age of 70 years and a decreased diffusing capacity of carbon monoxide. Ventilatory responses to hypercapnia, hypoxia and exercise decrease in the elderly. Knowledge of changes in the respiratory system related to advancing age is a medical issue of great importance in order to distinguish the effects of aging from those of diseases. Copyright © 2012 Elsevier Masson SAS. All rights reserved.

  8. Interface of RETRAN/MASTER Code System for APR1400

    International Nuclear Information System (INIS)

    Ku, Keuk Jong; Kang, Sang Hee; Kim, Han Gon

    2008-01-01

    MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), which was developed by KAERI, is a nuclear analysis and design code which can simulate the pressurized water reactor core or boiling water reactor core in 3-dimensional geometry. RETRAN is a best-estimate code for transient analysis of Non-LOCA. RETRAN code generates neutron number density in core using point kinetics model which includes feedback reactivities and converts the neutron number density into reactor power. It is conventional that RETRAN code for power generation is roughly to extrapolate feedback reactivities which are provided by MASTER code only one time before transient analysis. The purpose of this paper is to interface RETRAN code with MASTER code by real-time processing and to supply adequate feedback reactivities to RETRAN code. So, we develop interface code called MATRAN for real-time feedback reactivity processing. And for the application of MATRAN code, we compare the results of real-time MATRAN code with those of conventional RETRAN/MASTER code

  9. Implications of Sepedi/English code switching for ASR systems

    CSIR Research Space (South Africa)

    Modipa, TI

    2013-12-01

    Full Text Available . We also perform an initial acoustic analysis to determine the impact of such code switching on speech recognition performance. We nd that the frequency of code switching is unexpectedly high, and that the continuum of code switching (from unmodi ed...

  10. Defending public interests in private lands: compliance, costs and potential environmental consequences of the Brazilian Forest Code in Mato Grosso.

    Science.gov (United States)

    Stickler, Claudia M; Nepstad, Daniel C; Azevedo, Andrea A; McGrath, David G

    2013-06-05

    Land-use regulations are a critical component of forest governance and conservation strategies, but their effectiveness in shaping landholder behaviour is poorly understood. We conducted a spatial and temporal analysis of the Brazilian Forest Code (BFC) to understand the patterns of regulatory compliance over time and across changes in the policy, and the implications of these compliance patterns for the perceived costs to landholders and environmental performance of agricultural landscapes in the southern Amazon state of Mato Grosso. Landholdings tended to remain in compliance or not according to their status at the beginning of the study period. The perceived economic burden of BFC compliance on soya bean and beef producers (US$3-5.6 billion in net present value of the land) may in part explain the massive, successful campaign launched by the farm lobby to change the BFC. The ecological benefits of compliance (e.g. greater connectivity and carbon) with the BFC are diffuse and do not compete effectively with the economic benefits of non-compliance that are perceived by landholders. Volatile regulation of land-use decisions that affect billions in economic rent that could be captured is an inadequate forest governance instrument; effectiveness of such regulations may increase when implemented in tandem with positive incentives for forest conservation.

  11. Simulation realization of 2-D wavelength/time system utilizing MDW code for OCDMA system

    Science.gov (United States)

    Azura, M. S. A.; Rashidi, C. B. M.; Aljunid, S. A.; Endut, R.; Ali, N.

    2017-11-01

    This paper presents a realization of Wavelength/Time (W/T) Two-Dimensional Modified Double Weight (2-D MDW) code for Optical Code Division Multiple Access (OCDMA) system based on Spectral Amplitude Coding (SAC) approach. The MDW code has the capability to suppress Phase-Induce Intensity Noise (PIIN) and minimizing the Multiple Access Interference (MAI) noises. At the permissible BER 10-9, the 2-D MDW (APD) had shown minimum effective received power (Psr) = -71 dBm that can be obtained at the receiver side as compared to 2-D MDW (PIN) only received -61 dBm. The results show that 2-D MDW (APD) has better performance in achieving same BER with longer optical fiber length and with less received power (Psr). Also, the BER from the result shows that MDW code has the capability to suppress PIIN ad MAI.

  12. Simulation realization of 2-D wavelength/time system utilizing MDW code for OCDMA system

    Directory of Open Access Journals (Sweden)

    Azura M. S. A.

    2017-01-01

    Full Text Available This paper presents a realization of Wavelength/Time (W/T Two-Dimensional Modified Double Weight (2-D MDW code for Optical Code Division Multiple Access (OCDMA system based on Spectral Amplitude Coding (SAC approach. The MDW code has the capability to suppress Phase-Induce Intensity Noise (PIIN and minimizing the Multiple Access Interference (MAI noises. At the permissible BER 10-9, the 2-D MDW (APD had shown minimum effective received power (Psr = -71 dBm that can be obtained at the receiver side as compared to 2-D MDW (PIN only received -61 dBm. The results show that 2-D MDW (APD has better performance in achieving same BER with longer optical fiber length and with less received power (Psr. Also, the BER from the result shows that MDW code has the capability to suppress PIIN ad MAI.

  13. Software coding for reliable data communication in a reactor safety system

    International Nuclear Information System (INIS)

    Maghsoodi, R.

    1978-01-01

    A software coding method is proposed to improve the communication reliability of a microprocessor based fast-reactor safety system. This method which replaces the conventional coding circuitry, applies a program to code the data which is communicated between the processors via their data memories. The system requirements are studied and the suitable codes are suggested. The problems associated with hardware coders, and the advantages of software coding methods are discussed. The product code which proves a faster coding time over the cyclic code is chosen as the final code. Then the improvement of the communication reliability is derived for a processor and its data memory. The result is used to calculate the reliability improvement of the processing channel as the basic unit for the safety system. (author)

  14. AUS98 - The 1998 version of the AUS modular neutronic code system

    International Nuclear Information System (INIS)

    Robinson, G.S.; Harrington, B.V.

    1998-07-01

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module

  15. AUS98 - The 1998 version of the AUS modular neutronic code system

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, G.S.; Harrington, B.V

    1998-07-01

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module refs., tabs.

  16. Evaluation and implementation of QR Code Identity Tag system for Healthcare in Turkey

    OpenAIRE

    Uzun, Vassilya; Bilgin, Sami

    2016-01-01

    For this study, we designed a QR Code Identity Tag system to integrate into the Turkish healthcare system. This system provides QR code-based medical identification alerts and an in-hospital patient identification system. Every member of the medical system is assigned a unique QR Code Tag; to facilitate medical identification alerts, the QR Code Identity Tag can be worn as a bracelet or necklace or carried as an ID card. Patients must always possess the QR Code Identity bracelets within hospi...

  17. Development of an accident consequence assessment code for evaluating site suitability of light- and heavy-water reactors based on the Korean Technical standards

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Won Tae; Jeong, Hae Sung; Jeong, Hyo Joon; Kil, A Reum; Kim, Eun Han; Han, Moon Hee [Nuclear Environment Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

  18. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  19. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

    2011-01-01

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  20. Pathology of motivation system – reasons and consequences

    Directory of Open Access Journals (Sweden)

    Ewa Mazanowska

    2013-09-01

    Full Text Available The aim of this study is to define the dysfunction of the motivational process and the presentation accompanying this issue causes and effects. Pathologies of the incentive system in this day and age are quite a frequently occurring phenomenon in Polish companies. This is accompanied by increasing ignorance of employers in the area of motivation, incompetence immediate superiors in the field of personnel management, as well as advancing the level of unemployment. Employees are evaluated through the prism of cheap labour and so they are treated. Lack of motivation and appropriate way to combat forms of harassment or discrimination in the workplace causes subordinates loss of self-esteem and commitment to duties performed. Thus, this could lead to excessive stress or so burnout, and sometimes even the death of the employee.

  1. Verification of the CONPAS (CONtainment Performance Analysis System) code package

    International Nuclear Information System (INIS)

    Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

    1997-09-01

    CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

  2. Pressure vessel codes: Their application to nuclear reactor systems

    International Nuclear Information System (INIS)

    1966-01-01

    A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

  3. The APR1400 Core Design by Using APA Code System

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Koh, Byung Marn

    2008-01-01

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

  4. Biometric iris image acquisition system with wavefront coding technology

    Science.gov (United States)

    Hsieh, Sheng-Hsun; Yang, Hsi-Wen; Huang, Shao-Hung; Li, Yung-Hui; Tien, Chung-Hao

    2013-09-01

    Biometric signatures for identity recognition have been practiced for centuries. Basically, the personal attributes used for a biometric identification system can be classified into two areas: one is based on physiological attributes, such as DNA, facial features, retinal vasculature, fingerprint, hand geometry, iris texture and so on; the other scenario is dependent on the individual behavioral attributes, such as signature, keystroke, voice and gait style. Among these features, iris recognition is one of the most attractive approaches due to its nature of randomness, texture stability over a life time, high entropy density and non-invasive acquisition. While the performance of iris recognition on high quality image is well investigated, not too many studies addressed that how iris recognition performs subject to non-ideal image data, especially when the data is acquired in challenging conditions, such as long working distance, dynamical movement of subjects, uncontrolled illumination conditions and so on. There are three main contributions in this paper. Firstly, the optical system parameters, such as magnification and field of view, was optimally designed through the first-order optics. Secondly, the irradiance constraints was derived by optical conservation theorem. Through the relationship between the subject and the detector, we could estimate the limitation of working distance when the camera lens and CCD sensor were known. The working distance is set to 3m in our system with pupil diameter 86mm and CCD irradiance 0.3mW/cm2. Finally, We employed a hybrid scheme combining eye tracking with pan and tilt system, wavefront coding technology, filter optimization and post signal recognition to implement a robust iris recognition system in dynamic operation. The blurred image was restored to ensure recognition accuracy over 3m working distance with 400mm focal length and aperture F/6.3 optics. The simulation result as well as experiment validates the proposed code

  5. Deterministic sensitivity and uncertainty methodology for best estimate system codes applied in nuclear technology

    International Nuclear Information System (INIS)

    Petruzzi, A.; D'Auria, F.; Cacuci, D.G.

    2009-01-01

    Nuclear Power Plant (NPP) technology has been developed based on the traditional defense in depth philosophy supported by deterministic and overly conservative methods for safety analysis. In the 1970s [1], conservative hypotheses were introduced for safety analyses to address existing uncertainties. Since then, intensive thermal-hydraulic experimental research has resulted in a considerable increase in knowledge and consequently in the development of best-estimate codes able to provide more realistic information about the physical behaviour and to identify the most relevant safety issues allowing the evaluation of the existing actual margins between the results of the calculations and the acceptance criteria. However, the best-estimate calculation results from complex thermal-hydraulic system codes (like Relap5, Cathare, Athlet, Trace, etc..) are affected by unavoidable approximations that are un-predictable without the use of computational tools that account for the various sources of uncertainty. Therefore the use of best-estimate codes (BE) within the reactor technology, either for design or safety purposes, implies understanding and accepting the limitations and the deficiencies of those codes. Taking into consideration the above framework, a comprehensive approach for utilizing quantified uncertainties arising from Integral Test Facilities (ITFs, [2]) and Separate Effect Test Facilities (SETFs, [3]) in the process of calibrating complex computer models for the application to NPP transient scenarios has been developed. The methodology proposed is capable of accommodating multiple SETFs and ITFs to learn as much as possible about uncertain parameters, allowing for the improvement of the computer model predictions based on the available experimental evidences. The proposed methodology constitutes a major step forward with respect to the generally used expert judgment and statistical methods as it permits a) to establish the uncertainties of any parameter

  6. RFSYS: an inventory code for RF system parameters

    International Nuclear Information System (INIS)

    Treadwell, E.A.

    1983-03-01

    RFSYS is a program which maintains an inventory of rf system parameters associated with the 200 MeV Linear Accelerator at Fermi National Accelerator Laboratory. The program, written by Elliott Treadwell, of the Linac group, offers five modes of operation: (1) Allocates memory space for additional rf systems (data arrays). (2) Prints a total or partial list of old tube parameters on an ADM-3 terminal. (3) Changes tube data stored in the master array. If the number of systems increases, this mode permits the user to enter new data. (4) Computes the average time of operation for a given tube and system. (5) Stops program execution. There is an exit option, (a) create one output data file or (b) create three output files, one of which contains column headers and coded comments. All output files are stored on the CYBER-175 disc, and eventually on high density (6250 B.P.I.) magnetic tapes. This arrangement eliminates the necessity for online data buffers

  7. Control code for laboratory adaptive optics teaching system

    Science.gov (United States)

    Jin, Moonseob; Luder, Ryan; Sanchez, Lucas; Hart, Michael

    2017-09-01

    By sensing and compensating wavefront aberration, adaptive optics (AO) systems have proven themselves crucial in large astronomical telescopes, retinal imaging, and holographic coherent imaging. Commercial AO systems for laboratory use are now available in the market. One such is the ThorLabs AO kit built around a Boston Micromachines deformable mirror. However, there are limitations in applying these systems to research and pedagogical projects since the software is written with limited flexibility. In this paper, we describe a MATLAB-based software suite to interface with the ThorLabs AO kit by using the MATLAB Engine API and Visual Studio. The software is designed to offer complete access to the wavefront sensor data, through the various levels of processing, to the command signals to the deformable mirror and fast steering mirror. In this way, through a MATLAB GUI, an operator can experiment with every aspect of the AO system's functioning. This is particularly valuable for tests of new control algorithms as well as to support student engagement in an academic environment. We plan to make the code freely available to the community.

  8. Modelling guidelines for core exit temperature simulations with system codes

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

    2015-05-15

    Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

  9. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  10. Performance analysis of multiple interference suppression over asynchronous/synchronous optical code-division multiple-access system based on complementary/prime/shifted coding scheme

    Science.gov (United States)

    Nieh, Ta-Chun; Yang, Chao-Chin; Huang, Jen-Fa

    2011-08-01

    A complete complementary/prime/shifted prime (CPS) code family for the optical code-division multiple-access (OCDMA) system is proposed. Based on the ability of complete complementary (CC) code, the multiple-access interference (MAI) can be suppressed and eliminated via spectral amplitude coding (SAC) OCDMA system under asynchronous/synchronous transmission. By utilizing the shifted prime (SP) code in the SAC scheme, the hardware implementation of encoder/decoder can be simplified with a reduced number of optical components, such as arrayed waveguide grating (AWG) and fiber Bragg grating (FBG). This system has a superior performance as compared to previous bipolar-bipolar coding OCDMA systems.

  11. A study on the nuclear computer codes installation and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Kim, Hee Kyung; Kang, Byung Heon; Kim, Ko Ryeo; Suh, Soong Hyok; Choi, Young Gil; Lee, Jong Bok

    1990-12-01

    From 1987 a number of technical transfer related to nuclear power plant had been performed from C-E for YGN 3 and 4 construction. Among them, installation and management of the computer codes for YGN 3 and 4 fuel and nuclear steam supply system was one of the most important project. Main objectives of this project are to establish the nuclear computer code management system, to develop QA procedure for nuclear codes, to secure the nuclear code reliability and to extend techanical applicabilities including the user-oriented utility programs for nuclear codes. Contents of performing the project in this year was to produce 215 transmittal packages of nuclear codes installation including making backup magnetic tape and microfiche for software quality assurance. Lastly, for easy reference about the nuclear codes information we presented list of code names and information on the codes which were introduced from C-E. (Author)

  12. Performance Analysis of Wavelength Multiplexed Sac Ocdma Codes in Beat Noise Mitigation in Sac Ocdma Systems

    Science.gov (United States)

    Alhassan, A. M.; Badruddin, N.; Saad, N. M.; Aljunid, S. A.

    2013-07-01

    In this paper we investigate the use of wavelength multiplexed spectral amplitude coding (WM SAC) codes in beat noise mitigation in coherent source SAC OCDMA systems. A WM SAC code is a low weight SAC code, where the whole code structure is repeated diagonally (once or more) in the wavelength domain to achieve the same cardinality as a higher weight SAC code. Results show that for highly populated networks, the WM SAC codes provide better performance than SAC codes. However, for small number of active users the situation is reversed. Apart from their promising improvement in performance, these codes are more flexible and impose less complexity on the system design than their SAC counterparts.

  13. Comparison of PSF maxima and minima of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems

    Science.gov (United States)

    Ratnam, Challa; Lakshmana Rao, Vadlamudi; Lachaa Goud, Sivagouni

    2006-10-01

    In the present paper, and a series of papers to follow, the Fourier analytical properties of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems are investigated. First, the transmission function for MACA and CMACA is derived using Fourier methods and, based on the Fresnel-Kirchoff diffraction theory, the formulae for the point spread function are formulated. The PSF maxima and minima are calculated for both the MACA and CMACA systems. The dependence of these properties on the number of zones is studied and reported in this paper.

  14. Comparison of PSF maxima and minima of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems

    International Nuclear Information System (INIS)

    Ratnam, Challa; Rao, Vadlamudi Lakshmana; Goud, Sivagouni Lachaa

    2006-01-01

    In the present paper, and a series of papers to follow, the Fourier analytical properties of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems are investigated. First, the transmission function for MACA and CMACA is derived using Fourier methods and, based on the Fresnel-Kirchoff diffraction theory, the formulae for the point spread function are formulated. The PSF maxima and minima are calculated for both the MACA and CMACA systems. The dependence of these properties on the number of zones is studied and reported in this paper

  15. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  16. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-08-15

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  17. Novel BCH Code Design for Mitigation of Phase Noise Induced Cycle Slips in DQPSK Systems

    DEFF Research Database (Denmark)

    Leong, M. Y.; Larsen, Knud J.; Jacobsen, G.

    2014-01-01

    We show that by proper code design, phase noise induced cycle slips causing an error floor can be mitigated for 28 Gbau d DQPSK systems. Performance of BCH codes are investigated in terms of required overhead......We show that by proper code design, phase noise induced cycle slips causing an error floor can be mitigated for 28 Gbau d DQPSK systems. Performance of BCH codes are investigated in terms of required overhead...

  18. Variable weight Khazani-Syed code using hybrid fixed-dynamic technique for optical code division multiple access system

    Science.gov (United States)

    Anas, Siti Barirah Ahmad; Seyedzadeh, Saleh; Mokhtar, Makhfudzah; Sahbudin, Ratna Kalos Zakiah

    2016-10-01

    Future Internet consists of a wide spectrum of applications with different bit rates and quality of service (QoS) requirements. Prioritizing the services is essential to ensure that the delivery of information is at its best. Existing technologies have demonstrated how service differentiation techniques can be implemented in optical networks using data link and network layer operations. However, a physical layer approach can further improve system performance at a prescribed received signal quality by applying control at the bit level. This paper proposes a coding algorithm to support optical domain service differentiation using spectral amplitude coding techniques within an optical code division multiple access (OCDMA) scenario. A particular user or service has a varying weight applied to obtain the desired signal quality. The properties of the new code are compared with other OCDMA codes proposed for service differentiation. In addition, a mathematical model is developed for performance evaluation of the proposed code using two different detection techniques, namely direct decoding and complementary subtraction.

  19. Multilevel LDPC Codes Design for Multimedia Communication CDMA System

    Directory of Open Access Journals (Sweden)

    Hou Jia

    2004-01-01

    Full Text Available We design multilevel coding (MLC with a semi-bit interleaved coded modulation (BICM scheme based on low density parity check (LDPC codes. Different from the traditional designs, we joined the MLC and BICM together by using the Gray mapping, which is suitable to transmit the data over several equivalent channels with different code rates. To perform well at signal-to-noise ratio (SNR to be very close to the capacity of the additive white Gaussian noise (AWGN channel, random regular LDPC code and a simple semialgebra LDPC (SA-LDPC code are discussed in MLC with parallel independent decoding (PID. The numerical results demonstrate that the proposed scheme could achieve both power and bandwidth efficiency.

  20. Development of System Based Code: Case Study of Life-Cycle Margin Evaluation

    International Nuclear Information System (INIS)

    Tai Asayama; Masaki Morishita; Masanori Tashimo

    2006-01-01

    For a leap of progress in structural deign of nuclear plant components, The late Professor Emeritus Yasuhide Asada proposed the System Based Code. The key concepts of the System Based Code are; (1) life-cycle margin optimization, (2) expansion of technical options as well as combinations of technical options beyond the current codes and standards, and (3) designing to clearly defined target reliabilities. Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to. On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept. It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code. (authors)

  1. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  2. Application of GIS in prediction and assessment system of off-site accident consequence for NPP

    International Nuclear Information System (INIS)

    Wang Xingyu; Shi Zhongqi

    2002-01-01

    The assessment and prediction software system of off-site accident consequence for Guangdong Nuclear Power Plant (GNARD2.0) is a GIS-based software system. The spatial analysis of radioactive materials and doses with geographic information is available in this system. The structure and functions of the GNARD system and the method of applying ArcView GIS are presented

  3. Prototype demonstration of radiation therapy planning code system

    International Nuclear Information System (INIS)

    Little, R.C.; Adams, K.J.; Estes, G.P.; Hughes, L.S. III; Waters, L.S.

    1996-01-01

    This is the final report of a one-year, Laboratory-Directed Research and Development project at the Los Alamos National Laboratory (LANL). Radiation therapy planning is the process by which a radiation oncologist plans a treatment protocol for a patient preparing to undergo radiation therapy. The objective is to develop a protocol that delivers sufficient radiation dose to the entire tumor volume, while minimizing dose to healthy tissue. Radiation therapy planning, as currently practiced in the field, suffers from inaccuracies made in modeling patient anatomy and radiation transport. This project investigated the ability to automatically model patient-specific, three-dimensional (3-D) geometries in advanced Los Alamos radiation transport codes (such as MCNP), and to efficiently generate accurate radiation dose profiles in these geometries via sophisticated physics modeling. Modem scientific visualization techniques were utilized. The long-term goal is that such a system could be used by a non-expert in a distributed computing environment to help plan the treatment protocol for any candidate radiation source. The improved accuracy offered by such a system promises increased efficacy and reduced costs for this important aspect of health care

  4. Web- and system-code based, interactive, nuclear power plant simulators

    International Nuclear Information System (INIS)

    Kim, K. D.; Jain, P.; Rizwan, U.

    2006-01-01

    Using two different approaches, on-line, web- and system-code based graphical user interfaces have been developed for reactor system analysis. Both are LabVIEW (graphical programming language developed by National Instruments) based systems that allow local users as well as those at remote sites to run, interact and view the results of the system code in a web browser. In the first approach, only the data written by the system code in a tab separated ASCII output file is accessed and displayed graphically. In the second approach, LabVIEW virtual instruments are coupled with the system code as dynamic link libraries (DLL). RELAP5 is used as the system code to demonstrate the capabilities of these approaches. From collaborative projects between teams in geographically remote locations to providing system code experience to distance education students, these tools can be very beneficial in many areas of teaching and R and D. (authors)

  5. Coding the Assembly of Polyoxotungstates with a Programmable Reaction System.

    Science.gov (United States)

    Ruiz de la Oliva, Andreu; Sans, Victor; Miras, Haralampos N; Long, De-Liang; Cronin, Leroy

    2017-05-01

    Chemical transformations are normally conducted in batch or flow mode, thereby allowing the chemistry to be temporally or spatially controlled, but these approaches are not normally combined dynamically. However, the investigation of the underlying chemistry masked by the self-assembly processes that often occur in one-pot reactions and exploitation of the potential of complex chemical systems requires control in both time and space. Additionally, maintaining the intermediate constituents of a self-assembled system "off equilibrium" and utilizing them dynamically at specific time intervals provide access to building blocks that cannot coexist under one-pot conditions and ultimately to the formation of new clusters. Herein, we implement the concept of a programmable networked reaction system, allowing us to connect discrete "one-pot" reactions that produce the building block{W 11 O 38 } ≡ {W 11 } under different conditions and control, in real time, the assembly of a series of polyoxometalate clusters {W 12 O 42 } ≡ {W 12 }, {W 22 O 74 } ≡ {W 22 } 1a, {W 34 O 116 } ≡ {W 34 } 2a, and {W 36 O 120 } ≡ {W 36 } 3a, using pH and ultraviolet-visible monitoring. The programmable networked reaction system reveals that is possible to assemble a range of different clusters using {W 11 }-based building blocks, demonstrating the relationship between the clusters within the family of iso-polyoxotungstates, with the final structural motif being entirely dependent on the building block libraries generated in each separate reaction space within the network. In total, this approach led to the isolation of five distinct inorganic clusters using a "fixed" set of reagents and using a fully automated sequence code, rather than five entirely different reaction protocols. As such, this approach allows us to discover, record, and implement complex one-pot reaction syntheses in a more general way, increasing the yield and reproducibility and potentially giving access to

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

    International Nuclear Information System (INIS)

    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.

    2011-01-01

    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 5x5x50 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.

  7. A comparison of the consequences of different waste handling systems in two Danish communities

    DEFF Research Database (Denmark)

    Grunert, Suzanne C.; Thøgersen, John

    1995-01-01

    a system based solely on non-economic incentives. The main objective was to compare citizen`s beliefs and attitudes towards waste handling systems and their consequence for motivations to co-operate. Th groups of hypotheses concerning the beliefs-attitude relationship, differences in attitudes between...... cities, and the use of economic incentives were tested. Whereas beliefs influenced attitudes in the expected direction, the consequences of economi incentives for differences in attitudes were less clear....

  8. Evaluation and implementation of QR Code Identity Tag system for Healthcare in Turkey.

    Science.gov (United States)

    Uzun, Vassilya; Bilgin, Sami

    2016-01-01

    For this study, we designed a QR Code Identity Tag system to integrate into the Turkish healthcare system. This system provides QR code-based medical identification alerts and an in-hospital patient identification system. Every member of the medical system is assigned a unique QR Code Tag; to facilitate medical identification alerts, the QR Code Identity Tag can be worn as a bracelet or necklace or carried as an ID card. Patients must always possess the QR Code Identity bracelets within hospital grounds. These QR code bracelets link to the QR Code Identity website, where detailed information is stored; a smartphone or standalone QR code scanner can be used to scan the code. The design of this system allows authorized personnel (e.g., paramedics, firefighters, or police) to access more detailed patient information than the average smartphone user: emergency service professionals are authorized to access patient medical histories to improve the accuracy of medical treatment. In Istanbul, we tested the self-designed system with 174 participants. To analyze the QR Code Identity Tag system's usability, the participants completed the System Usability Scale questionnaire after using the system.

  9. Calculation code for erosion-corrosion induced wall thinning in piping systems

    International Nuclear Information System (INIS)

    Henzel, N.; Kastner, W.; Stellwag, B.; Erve, M.

    1988-01-01

    There was great material erosion mainly in consequence of an extremely unfavourable geometry at the damaged place in Surry-2. The pipeline sections affected in Trojan were in the area of action of great sources of turbulence, i.e.: less than 10 pipe diameters from junctions, elbows etc. Because of the many parameters which determine the amount of material removal by erosion-corrosion, the analysis of such damage is only possible using a computer program. The main purpose of such a PC code called WATHEC developed by Siemens/KWU is not the subsequent confirmation of damage which has occurred, but its application for preventive diagnosis in pipeline systems. (orig./DG) [de

  10. System Level Evaluation of Innovative Coded MIMO-OFDM Systems for Broadcasting Digital TV

    Directory of Open Access Journals (Sweden)

    Y. Nasser

    2008-01-01

    Full Text Available Single-frequency networks (SFNs for broadcasting digital TV is a topic of theoretical and practical interest for future broadcasting systems. Although progress has been made in the characterization of its description, there are still considerable gaps in its deployment with MIMO technique. The contribution of this paper is multifold. First, we investigate the possibility of applying a space-time (ST encoder between the antennas of two sites in SFN. Then, we introduce a 3D space-time-space block code for future terrestrial digital TV in SFN architecture. The proposed 3D code is based on a double-layer structure designed for intercell and intracell space time-coded transmissions. Eventually, we propose to adapt a technique called effective exponential signal-to-noise ratio (SNR mapping (EESM to predict the bit error rate (BER at the output of the channel decoder in the MIMO systems. The EESM technique as well as the simulations results will be used to doubly check the efficiency of our 3D code. This efficiency is obtained for equal and unequal received powers whatever is the location of the receiver by adequately combining ST codes. The 3D code is then a very promising candidate for SFN architecture with MIMO transmission.

  11. Development and verification of a coupled code system RETRAN-MASTER-TORC

    International Nuclear Information System (INIS)

    Cho, J.Y.; Song, J.S.; Joo, H.G.; Zee, S.Q.

    2004-01-01

    Recently, coupled thermal-hydraulics (T-H) and three-dimensional kinetics codes have been widely used for the best-estimate simulations such as the main steam line break (MSLB) and locked rotor problems. This work is to develop and verify one of such codes by coupling the system T-H code RETRAN, the 3-D kinetics code MASTER and sub-channel analysis code TORC. The MASTER code has already been applied to such simulations after coupling with the MARS or RETRAN-3D multi-dimensional system T-H codes. The MASTER code contains a sub-channel analysis code COBRA-III C/P, and the coupled systems MARSMASTER-COBRA and RETRAN-MASTER-COBRA had been already developed and verified. With these previous studies, a new coupled system of RETRAN-MASTER-TORC is to be developed and verified for the standard best-estimate simulation code package in Korea. The TORC code has already been applied to the thermal hydraulics design of the several ABB/CE type plants and Korean Standard Nuclear Power Plants (KSNP). This justifies the choice of TORC rather than COBRA. Because the coupling between RETRAN and MASTER codes are already established and verified, this work is simplified to couple the TORC sub-channel T-H code with the MASTER neutronics code. The TORC code is a standalone code that solves the T-H equations for a given core problem from reading the input file and finally printing the converged solutions. However, in the coupled system, because TORC receives the pin power distributions from the neutronics code MASTER and transfers the T-H results to MASTER iteratively, TORC needs to be controlled by the MASTER code and does not need to solve the given problem completely at each iteration step. By this reason, the coupling of the TORC code with the MASTER code requires several modifications in the I/O treatment, flow iteration and calculation logics. The next section of this paper describes the modifications in the TORC code. The TORC control logic of the MASTER code is then followed. The

  12. A Spanish version for the new ERA-EDTA coding system for primary renal disease

    Directory of Open Access Journals (Sweden)

    Óscar Zurriaga

    2015-07-01

    Conclusions: Translation and adaptation into Spanish represent an improvement that will help to introduce and use the new coding system for PKD, as it can help reducing the time devoted to coding and also the period of adaptation of health workers to the new codes.

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

    International Nuclear Information System (INIS)

    Paratte, J.M.; Grimm, P.; Hollard, J.M.

    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

  14. Performance analysis of wavelength/spatial coding system with fixed in-phase code matrices in OCDMA network

    Science.gov (United States)

    Tsai, Cheng-Mu; Liang, Tsair-Chun

    2011-12-01

    This paper proposes a wavelength/spatial (W/S) coding system with fixed in-phase code (FIPC) matrix in the optical code-division multiple-access (OCDMA) network. A scheme is presented to form the FIPC matrix which is applied to construct the W/S OCDMA network. The encoder/decoder in the W/S OCDMA network is fully able to eliminate the multiple-access-interference (MAI) at the balanced photo-detectors (PD), according to fixed in-phase cross correlation. The phase-induced intensity noise (PIIN) related to the power square is markedly suppressed in the receiver by spreading the received power into each PD while the net signal power is kept the same. Simulation results show that the W/S OCDMA network based on the FIPC matrices cannot only completely remove the MAI but effectively suppress the PIIN to upgrade the network performance.

  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).

  16. Updated Covariance Processing Capabilities in the AMPX Code System

    International Nuclear Information System (INIS)

    Wiarda, Dorothea; Dunn, Michael E.

    2007-01-01

    A concerted effort is in progress within the nuclear data community to provide new cross-section covariance data evaluations to support sensitivity/uncertainty analyses of fissionable systems. The objective of this work is to update processing capabilities of the AMPX library to process the latest Evaluated Nuclear Data File (ENDF)/B formats to generate covariance data libraries for radiation transport software such as SCALE. The module PUFF-IV was updated to allow processing of new ENDF covariance formats in the resolved resonance region. In the resolved resonance region, covariance matrices are given in terms of resonance parameters, which need to be processed into covariance matrices with respect to the group-averaged cross-section data. The parameter covariance matrix can be quite large if the evaluation has many resonances. The PUFF-IV code has recently been used to process an evaluation of 235U, which was prepared in collaboration between Oak Ridge National Laboratory and Los Alamos National Laboratory.

  17. 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.

  18. Classification Systems, their Digitization and Consequences for Data-Driven Decision Making

    DEFF Research Database (Denmark)

    Stein, Mari-Klara; Newell, Sue; Galliers, Robert D.

    2013-01-01

    Classification systems are foundational in many standardized software tools. This digitization of classification systems gives them a new ‘materiality’ that, jointly with the social practices of information producers/consumers, has significant consequences on the representational quality of such ...... and the foundational role of representational quality in understanding the success and consequences of data-driven decision-making.......-narration and meta-narration), and three different information production/consumption situations. We contribute to the relational theorization of representational quality and extend classification systems research by drawing explicit attention to the importance of ‘materialization’ of classification systems...

  19. Library system for a one dimensional tokamak transport code: (LIBJT60), 1

    International Nuclear Information System (INIS)

    Hirayama, Toshio

    1982-12-01

    A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

  20. Radioactivities evaluation code system for high temperature gas cooled reactors during normal operation

    International Nuclear Information System (INIS)

    Ogura, Kenji; Morimoto, Toshio; Suzuki, Katsuo.

    1979-01-01

    A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)

  1. About the coding system of rivers, catchment basing and their characteristics of the republic of Armenia

    International Nuclear Information System (INIS)

    Avagyan, A.A.; Arakelyan, A.A.

    2011-01-01

    The coding of rivers, catchements, lakes and seas is one of the most important requirements of Water Framework Directive of the European Union. This coding provides solutions to actual problems of planning and management of water resources of the Republic of Armenia. The coding system provides the hierarchy of water bodies and watersheds with their typology as well as their geographic and natural conditions, anthropogenic pressures and ecological status. This approach is a fundamentally new complex solution to the coding of water resources. The coding technique allows you to automate the assessment and mapping of environmental risks and areas of water bodies which are subjected to significant pressure and also helps to solve other problems concerning the planning and the management of water resources. A complex code of each water body consists of the following groups of codes: Hydrographic code - an identifier of a water body in the hydrographic system of the country; Codes of static attributes in the system requirements of the Water Framework Directive of the European Union; Codes of static attributes of the qualifiers of the RA National Water Program; Codes of dynamic attributes that define the quality of water and characteristics of water use; Codes of dynamic attributes describing the human impact and determining the ecological status of water body

  2. Modular Modeling System (MMS) code: a versatile power plant analysis package

    International Nuclear Information System (INIS)

    Divakaruni, S.M.; Wong, F.K.L.

    1987-01-01

    The basic version of the Modular Modeling System (MMS-01), a power plant systems analysis computer code jointly developed by the Nuclear Power and the Coal Combustion Systems Divisions of the Electric Power Research Institute (EPRI), has been released to the utility power industry in April 1983 at a code release workshop held in Charlotte, North Carolina. Since then, additional modules have been developed to analyze the Pressurized Water Reactors (PWRs) and the Boiling Water Reactors (BWRs) when the safety systems are activated. Also, a selected number of modules in the MMS-01 library have been modified to allow the code users more flexibility in constructing plant specific systems for analysis. These new PWR and BWR modules constitute the new MMS library, and it includes the modifications to the MMS-01 library. A year and half long extensive code qualification program of this new version of the MMS code at EPRI and the contractor sites, back by further code testing in an user group environment is culminating in the MMS-02 code release announcement seminar. At this seminar, the results of user group efforts and the code qualification program will be presented in a series of technical sessions. A total of forty-nine papers will be presented to describe the new code features and the code qualification efforts. For the sake of completion, an overview of the code is presented to include the history of the code development, description of the MMS code and its structure, utility engineers involvement in MMS-01 and MMS-02 validations, the enhancements made in the last 18 months to the code, and finally the perspective on the code future in the fossil and nuclear industry

  3. 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.

  4. Modeling the consequences of hypothetical accidents for the Titan II system

    International Nuclear Information System (INIS)

    Greenly, G.D.; Sullivan, T.J.

    1981-11-01

    Calculations have been made with the Atmospheric Release Advisory Capability (ARAC) suite of three-dimensional transport and diffusion codes MATHEW/ADPIC to assess the consequences of severe, hypothetical accident scenarios. One set of calculations develops the integrated dose and surface deposition patterns for a non-nuclear, high explosive detonation and dispersal of material. A second set of calculations depicts the time integrated dose and instantaneous concentration patterns for a substantial, continuous leak of the missile fuel oxidizer converted to nitrogen dioxide (NO 2 ). The areas affected and some of the implications for emergency response management are discussed

  5. A coupled systems code-CFD MHD solver for fusion blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

    2015-10-15

    Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

  6. THYDE-P2 code: RCS (reactor-coolant system) analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo

    1986-12-01

    THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of through calculation from its initiation to complete reflooding of the core without an artificial change in the methods and models. The first half of the report is the description of the methods and models for use in the THYDE-P2 code, i.e., (1) the thermal-hydraulic network model, (2) the various RCS components models, (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 user's mannual for the THYDE-P2 code (version SV04L08A) containing items; (1) the program control (2) the input requirements, (3) the execution of THYDE-P2 job, (4) the output specifications and (5) the sample problem to demonstrate capability of the thermal-hydraulic network model, among other things. (author)

  7. Experimental transport analysis code system in JT-60

    International Nuclear Information System (INIS)

    Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

    1988-03-01

    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)

  8. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  9. Application of the cause-consequence diagram method to static systems

    International Nuclear Information System (INIS)

    Andrews, J.D.; Ridley, L.M.

    2002-01-01

    In the last 30 years, various mathematical models have been used to identify the effect of component failures on the performance of a system. The most frequently used technique for system reliability assessment is Fault Tree Analysis (FTA) and a large proportion of its popularity can be attributed to the fact that it provides a very good documentation of the way that the system failure logic was developed. Exact quantification of the fault tree, however, can be problematic for very large systems and in such situations, approximations can be used. Alternatively, an exact result can be obtained via the conversion of the fault tree into a binary decision diagram (BDD). The BDD, however, loses all failure logic documentation during the conversion process. This paper outlines the use of the cause-consequence diagram method as a tool for system risk and reliability analysis. As with the FTA method, the cause-consequence diagram documents the failure logic of the system. In addition to this the cause-consequence diagram produces the exact failure probability in a very efficient calculation procedure. The cause-consequence diagram technique has been applied to a static system and shown to yield the same result as those produced by the solution of the equivalent fault tree and BDD. On the basis of this general rules have been devised for the correct construction of the cause-consequence diagram given a static system. The use of the cause-consequence method in this manner has significant implications in terms of efficiency of the reliability analysis and can be shown to have benefits for static systems

  10. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    International Nuclear Information System (INIS)

    Son, Han Seong; Song, Deok Yong; Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon

    2006-01-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors

  11. 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.

  12. 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.

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

    International Nuclear Information System (INIS)

    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

  14. Construction and performance analysis of variable-weight optical orthogonal codes for asynchronous OCDMA systems

    Science.gov (United States)

    Li, Chuan-qi; Yang, Meng-jie; Zhang, Xiu-rong; Chen, Mei-juan; He, Dong-dong; Fan, Qing-bin

    2014-07-01

    A construction scheme of variable-weight optical orthogonal codes (VW-OOCs) for asynchronous optical code division multiple access (OCDMA) system is proposed. According to the actual situation, the code family can be obtained by programming in Matlab with the given code weight and corresponding capacity. The formula of bit error rate (BER) is derived by taking account of the effects of shot noise, avalanche photodiode (APD) bulk, thermal noise and surface leakage currents. The OCDMA system with the VW-OOCs is designed and improved. The study shows that the VW-OOCs have excellent performance of BER. Despite of coming from the same code family or not, the codes with larger weight have lower BER compared with the other codes in the same conditions. By taking simulation, the conclusion is consistent with the analysis of BER in theory. And the ideal eye diagrams are obtained by the optical hard limiter.

  15. High-Speed Turbo-TCM-Coded Orthogonal Frequency-Division Multiplexing Ultra-Wideband Systems

    Directory of Open Access Journals (Sweden)

    Wang Yanxia

    2006-01-01

    Full Text Available One of the UWB proposals in the IEEE P802.15 WPAN project is to use a multiband orthogonal frequency-division multiplexing (OFDM system and punctured convolutional codes for UWB channels supporting a data rate up to 480 Mbps. In this paper, we improve the proposed system using turbo TCM with QAM constellation for higher data rate transmission. We construct a punctured parity-concatenated trellis codes, in which a TCM code is used as the inner code and a simple parity-check code is employed as the outer code. The result shows that the system can offer a much higher spectral efficiency, for example, 1.2 Gbps, which is 2.5 times higher than the proposed system. We identify several essential requirements to achieve the high rate transmission, for example, frequency and time diversity and multilevel error protection. Results are confirmed by density evolution.

  16. The research of atmospheric 2D optical PPM CDMA system with turbo coding

    Science.gov (United States)

    Zhou, Xiuli; Li, Zaoxia

    2007-11-01

    The atmospheric two-dimensional optical code-division multiple-access (CDMA) systems using pulse-position modulation (PPM) and Turbo-coded were presented. We analyzed the bit-error rate (BER) of the proposed system using pulse-position modulation (PPM) with considering the effects of the scintillation, avalanche photodiode noise, thermal noise, and multi-user interference. We showed that the atmospheric two dimensional (2D) optical PPM CDMA systems can realize high-speed communications when the logarithm variance of the scintillation is less than 0.1, and the turbo-coded atmospheric optical CDMA system has better bit error rate(BER) performance than the atmospheric optical PPM CDMA systems without turbo-coded. We also showed that the turbo-coded system has better performance than the multi-user detection system.

  17. Code for calculation of spreading of radioactivity in reactor containment systems

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

    A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs

  18. Improvement of MSLB transient analysis for VVER by the coupled code system KIKO3D/ATHLET

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2001-01-01

    An overview is given on the investigations of the Main Steam Line Break transient in a VVER- 440 NPP by using the KIKO3D/ATHLET 1.2.A coupled code system. Special attention was paid for the influence of modeling the outcore detector signals and the malfunctioning of the emergency control system (scram with stuck rod). The conservatism of the calculations was assured even in the case of application of the 3D best estimate KIKO3D code. The consequence of MSLB accident is investigated at the end of cycle (EOC), at full power (FP) and shut down initial conditions. Even if very strong conservative assumptions were applied, dangerous hot spots were not found in the supposed scenarios.(author)

  19. Channel coding for underwater acoustic single-carrier CDMA communication system

    Science.gov (United States)

    Liu, Lanjun; Zhang, Yonglei; Zhang, Pengcheng; Zhou, Lin; Niu, Jiong

    2017-01-01

    CDMA is an effective multiple access protocol for underwater acoustic networks, and channel coding can effectively reduce the bit error rate (BER) of the underwater acoustic communication system. For the requirements of underwater acoustic mobile networks based on CDMA, an underwater acoustic single-carrier CDMA communication system (UWA/SCCDMA) based on the direct-sequence spread spectrum is proposed, and its channel coding scheme is studied based on convolution, RA, Turbo and LDPC coding respectively. The implementation steps of the Viterbi algorithm of convolutional coding, BP and minimum sum algorithms of RA coding, Log-MAP and SOVA algorithms of Turbo coding, and sum-product algorithm of LDPC coding are given. An UWA/SCCDMA simulation system based on Matlab is designed. Simulation results show that the UWA/SCCDMA based on RA, Turbo and LDPC coding have good performance such that the communication BER is all less than 10-6 in the underwater acoustic channel with low signal to noise ratio (SNR) from -12 dB to -10dB, which is about 2 orders of magnitude lower than that of the convolutional coding. The system based on Turbo coding with Log-MAP algorithm has the best performance.

  20. Rotated Walsh-Hadamard Spreading with Robust Channel Estimation for a Coded MC-CDMA System

    Directory of Open Access Journals (Sweden)

    Raulefs Ronald

    2004-01-01

    Full Text Available We investigate rotated Walsh-Hadamard spreading matrices for a broadband MC-CDMA system with robust channel estimation in the synchronous downlink. The similarities between rotated spreading and signal space diversity are outlined. In a multiuser MC-CDMA system, possible performance improvements are based on the chosen detector, the channel code, and its Hamming distance. By applying rotated spreading in comparison to a standard Walsh-Hadamard spreading code, a higher throughput can be achieved. As combining the channel code and the spreading code forms a concatenated code, the overall minimum Hamming distance of the concatenated code increases. This asymptotically results in an improvement of the bit error rate for high signal-to-noise ratio. Higher convolutional channel code rates are mostly generated by puncturing good low-rate channel codes. The overall Hamming distance decreases significantly for the punctured channel codes. Higher channel code rates are favorable for MC-CDMA, as MC-CDMA utilizes diversity more efficiently compared to pure OFDMA. The application of rotated spreading in an MC-CDMA system allows exploiting diversity even further. We demonstrate that the rotated spreading gain is still present for a robust pilot-aided channel estimator. In a well-designed system, rotated spreading extends the performance by using a maximum likelihood detector with robust channel estimation at the receiver by about 1 dB.

  1. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  2. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  3. LHC beam dump system : analysis of beam commissioning, performance and the consequences of abnormal operation

    International Nuclear Information System (INIS)

    Kramer, T.

    2011-01-01

    The LHC accelerates proton beams to a momentum of up to 7 TeV/c. At this energy level and with nominal beam intensity the stored energy of 360 MJ per beam is sufficient to melt 500 kg of copper. In addition up to 10 GJ are stored within the LHC magnet system at top energy. lt is obvious that such a machine needs well designed safety and protection systems. The LHC Beam Dump System (LBDS) is such a system and one of the most critical once concerning machine protection and safe operation. lt is used to dispose of high intensity beams between 450 GeV and 7 TeV and is thus designed to fast extract beam in a loss free way and to transfer it to an external absorber. For each ring systems of 15 horizontal fast kicker magnets (MKD), 15 vertically deflecting magnetic septa (MSD) and 10 diluter kicker magnets (MKB) are installed. This thesis is concerned with the analysis of the LBDS performance under normal operating parameters as well as under abnormal conditions like in the event of asynchronous beam abort or missing MKD elements. Therefore a sophisticated simulation environment was developed based on the use of the MAD-X tracking code. A system of tracking jobs was set up to study failure cases and losses for various dump events. Those jobs can be distributed to available CPU power and be calculated in parallel. Studies into the consequences of abnormal beam dump actions have been performed. Different error scenarios have been evaluated including an asynchronous dump action, prefire cases, and the impact of different orbit and collimator settings. Losses at locations in the ring and the beam dump transfer lines have been quantified as a function of different settings of the dump system protection elements. The implications for the setup and operation of these protection elements are discussed. Particle distributions can be created according to the used orbit. Simulations with different orbit parameters (including magnet field errors, beam position read out errors

  4. Modeling of a planning system in radiotherapy and Nuclear Medicine using the MCNP6 code

    International Nuclear Information System (INIS)

    Massicano, Felipe

    2015-01-01

    Cancer therapy has many branches and one of them is the use of radiation sources as treatment leading method. Radiotherapy and nuclear medicine are examples of these treatment types. For using the ionization radiation as main tool for the therapy, there is the need of crafting many treatment simulation in order to maximum the tumoral tissue dose without surpass the dose limit in health tissue surrounding. Treatment planning systems (TPS) are systems which have the purpose of simulating these therapy types. Nuclear medicine and radiotherapy have many distinct features linked to the therapy mode and consequently they have different TPS destined for each. The radiotherapy TPS is more developed than the nuclear medicine TPS and by that reason the development of a TPS that was similar to the radiotherapy TPS, but enough generic for include other therapy types, it will contribute with significant advances in nuclear medicine and in others therapy types with radiation. Based on this, the goal of work was to model a TPS that utilizes the Monte Carlo N-Particle Transport code (MCNP6) in order to simulate radiotherapy therapy, nuclear medicine therapy and with potential for simulating other therapy types too. The result of this work was the creation of a Framework in Java language, object oriented, named IBMC which will assist in the development of new TPS with MCNP6 code. The IBMC allowed to develop rapidly and easily TPS for radiotherapy and nuclear medicine and the results were validated with systems already consolidated. The IBMC showed high potential for developing TPS by new therapy types. (author)

  5. LDPC concatenated space-time block coded system in multipath fading environment: Analysis and evaluation

    Directory of Open Access Journals (Sweden)

    Surbhi Sharma

    2011-06-01

    Full Text Available Irregular low-density parity-check (LDPC codes have been found to show exceptionally good performance for single antenna systems over a wide class of channels. In this paper, the performance of LDPC codes with multiple antenna systems is investigated in flat Rayleigh and Rician fading channels for different modulation schemes. The focus of attention is mainly on the concatenation of irregular LDPC codes with complex orthogonal space-time codes. Iterative decoding is carried out with a density evolution method that sets a threshold above which the code performs well. For the proposed concatenated system, the simulation results show that the QAM technique achieves a higher coding gain of 8.8 dB and 3.2 dB over the QPSK technique in Rician (LOS and Rayleigh (NLOS faded environments respectively.

  6. Coding hazardous tree failures for a data management system

    Science.gov (United States)

    Lee A. Paine

    1978-01-01

    Codes for automatic data processing (ADP) are provided for hazardous tree failure data submitted on Report of Tree Failure forms. Definitions of data items and suggestions for interpreting ambiguously worded reports are also included. The manual is intended to insure the production of accurate and consistent punched ADP cards which are used in transfer of the data to...

  7. Energy Efficient Error-Correcting Coding for Wireless Systems

    NARCIS (Netherlands)

    Shao, X.

    2010-01-01

    The wireless channel is a hostile environment. The transmitted signal does not only suffers multi-path fading but also noise and interference from other users of the wireless channel. That causes unreliable communications. To achieve high-quality communications, error correcting coding is required

  8. Internal Corrosion Control of Water Supply Systems Code of Practice

    Science.gov (United States)

    This Code of Practice is part of a series of publications by the IWA Specialist Group on Metals and Related Substances in Drinking Water. It complements the following IWA Specialist Group publications: 1. Best Practice Guide on the Control of Lead in Drinking Water 2. Best Prac...

  9. Symbol Stream Combining in a Convolutionally Coded System

    Science.gov (United States)

    Mceliece, R. J.; Pollara, F.; Swanson, L.

    1985-01-01

    Symbol stream combining has been proposed as a method for arraying signals received at different antennas. If convolutional coding and Viterbi decoding are used, it is shown that a Viterbi decoder based on the proposed weighted sum of symbol streams yields maximum likelihood decisions.

  10. Proposing a Web-Based Tutorial System to Teach Malay Language Braille Code to the Sighted

    Science.gov (United States)

    Wah, Lee Lay; Keong, Foo Kok

    2010-01-01

    The "e-KodBrailleBM Tutorial System" is a web-based tutorial system which is specially designed to teach, facilitate and support the learning of Malay Language Braille Code to individuals who are sighted. The targeted group includes special education teachers, pre-service teachers, and parents. Learning Braille code involves memorisation…

  11. Coupling of 3D neutronics models with the system code ATHLET

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    1999-01-01

    The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)

  12. 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…

  13. A comparative evaluation of NDR and PSAR using the CASMO-3/MASTER code system

    International Nuclear Information System (INIS)

    Sim, Jeoung Hun; Kim, Han Gon

    2009-01-01

    In order to validate nuclear design data such as the nuclear design report (NDR) and data in preliminary (or final) safety analysis report (PSAR/FSAR) and to use data for the conceptual design of new plants, the CASMO-3/MASTER code system is selected as utility code. The nuclear design of OPR1000 and APR1400 is performed with the DIT/ROCS code system. In contrast with this design code system, the accuracy of CASMO- 3/MASTER code system has not been verified. Relatively little design data has been calculated by the CASMO-3/MASTER code system for OPR1000 and APR1400 and a bias system has not been developed yet. As such, validation of the performance of the CASMO- 3/MASTER code system is necessary. In order to validate the performance of the CASMO- 3/MASTER code system and to develop a calculation methodology, a comparative evaluation with NDR of Ulchin unit 4, cycle 1(U4C1) and the PSAR of Shinkori units 3 and 4 is carried out. The results of this evaluation are presented in this paper

  14. Management-retrieval code system of fission barrier parameter sub-library

    International Nuclear Information System (INIS)

    Zhang Limin; Su Zongdi; Ge Zhigang

    1995-01-01

    The fission barrier parameter (FBP) library, which is a sub-library of Chinese Evaluated Nuclear Parameter library (CENPL), stores various popular used fission barrier parameters from different historical period, and could retrieve the required fission barrier parameters by using the management retrieval code system of the FBP sub-library. The function, feature and operation instruction of the code system are described briefly

  15. Safety Analysis in Large Volume Vacuum Systems Like Tokamak: Experiments and Numerical Simulation to Analyze Vacuum Ruptures Consequences

    Directory of Open Access Journals (Sweden)

    A. Malizia

    2014-01-01

    Full Text Available The large volume vacuum systems are used in many industrial operations and research laboratories. Accidents in these systems should have a relevant economical and safety impact. A loss of vacuum accident (LOVA due to a failure of the main vacuum vessel can result in a fast pressurization of the vessel and consequent mobilization dispersion of hazardous internal material through the braches. It is clear that the influence of flow fields, consequence of accidents like LOVA, on dust resuspension is a key safety issue. In order to develop this analysis an experimental facility is been developed: STARDUST. This last facility has been used to improve the knowledge about LOVA to replicate a condition more similar to appropriate operative condition like to kamaks. By the experimental data the boundary conditions have been extrapolated to give the proper input for the 2D thermofluid-dynamics numerical simulations, developed by the commercial CFD numerical code. The benchmark of numerical simulation results with the experimental ones has been used to validate and tune the 2D thermofluid-dynamics numerical model that has been developed by the authors to replicate the LOVA conditions inside STARDUST. In present work, the facility, materials, numerical model, and relevant results will be presented.

  16. Application of startup/core management code system to YGN 3 startup testing

    International Nuclear Information System (INIS)

    Chi, Sung Goo; Hah, Yung Joon; Doo, Jin Yong; Kim, Dae Kyum

    1995-01-01

    YGN 3 is the first nuclear power plant in Korea to use the fixed incore detector system for startup testing and core management. The startup/core management code system was developed from existing ABB-C-E codes and applied for YGN 3 startup testing, especially for physics and CPC(Core Protection Calculator)/COLSS (Core Operating Limit Supervisory System) related testing. The startup/core management code system consists of startup codes which include the CEBASE, CECOR, CEFAST and CEDOPS, and startup data reduction codes which include FLOWRATE, COREPERF, CALMET, and VARTAV. These codes were implemented on an HP/Apollo model 9000 series 400 workstation at the YGN 3 site and successfully applied to startup testing and core management. The startup codes made a great contribution in upgrading the reliability of test results and reducing the test period by taking and analyzing core data automatically. The data reduction code saved the manpower and time for test data reduction and decreased the chance for error in the analysis. It is expected that this code system will make similar contributions for reducing the startup testing duration of YGN 4 and UCN3,4

  17. Geometrical Approach to the Grid System in the KOPEC Pilot Code

    International Nuclear Information System (INIS)

    Lee, E. J.; Park, C. E.; Lee, S. Y.

    2008-01-01

    KOPEC has been developing a pilot code to analyze two phase flow. The earlier version of the pilot code adopts the geometry with one-dimensional structured mesh system. As the pilot code is required to handle more complex geometries, a systematic geometrical approach to grid system has been introduced. Grid system can be classified as two types; structured grid system and unstructured grid system. The structured grid system is simple to apply but is less flexible than the other. The unstructured grid system is more complicated than the structured grid system. But it is more flexible to model the geometry. Therefore, two types of grid systems are utilized to allow code users simplicity as well as the flexibility

  18. Reducing BER of spectral-amplitude coding optical code-division multiple-access systems by single photodiode detection technique

    Science.gov (United States)

    Al-Khafaji, H. M. R.; Aljunid, S. A.; Amphawan, A.; Fadhil, H. A.; Safar, A. M.

    2013-03-01

    In this paper, we present a single photodiode detection (SPD) technique for spectral-amplitude coding optical code-division multiple-access (SAC-OCDMA) systems. The proposed technique eliminates both phase-induced intensity noise (PIIN) and multiple-access interference (MAI) in the optical domain. Analytical results show that for 35 simultaneous users transmitting at data rate of 622 Mbps, the bit-error rate (BER) = 1.4x10^-28 for SPD technique is much better compared to 9.3x10^-6 and 9.6x10^-3 for the modified-AND as well as the AND detection techniques, respectively. Moreover, we verified the improved performance afforded by the proposed technique using data transmission simulations.

  19. A bar-code reader for an alpha-beta automatic counting system - FAG

    International Nuclear Information System (INIS)

    Levinson, S.; Shemesh, Y.; Ankry, N.; Assido, H.; German, U.; Peled, O.

    1996-01-01

    A bar-code laser system for sample number reading was integrated into the FAG Alpha-Beta automatic counting system. The sample identification by means of an attached bar-code label enables unmistakable and reliable attribution of results to the counted sample. Installation of the bar-code reader system required several modifications: Mechanical changes in the automatic sample changer, design and production of new sample holders, modification of the sample planchettes, changes in the electronic system, update of the operating software of the system (authors)

  20. A bar-code reader for an alpha-beta automatic counting system - FAG

    Energy Technology Data Exchange (ETDEWEB)

    Levinson, S; Shemesh, Y; Ankry, N; Assido, H; German, U; Peled, O [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    A bar-code laser system for sample number reading was integrated into the FAG Alpha-Beta automatic counting system. The sample identification by means of an attached bar-code label enables unmistakable and reliable attribution of results to the counted sample. Installation of the bar-code reader system required several modifications: Mechanical changes in the automatic sample changer, design and production of new sample holders, modification of the sample planchettes, changes in the electronic system, update of the operating software of the system (authors).

  1. DANTSYS: A diffusion accelerated neutral particle transport code system

    International Nuclear Information System (INIS)

    Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O'Dell, R.D.; Walters, W.F.

    1995-06-01

    The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZΘ symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing

  2. DANTSYS: A diffusion accelerated neutral particle transport code system

    Energy Technology Data Exchange (ETDEWEB)

    Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.

    1995-06-01

    The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.

  3. A Systematic Review of Coding Systems Used in Pharmacoepidemiology and Database Research.

    Science.gov (United States)

    Chen, Yong; Zivkovic, Marko; Wang, Tongtong; Su, Su; Lee, Jianyi; Bortnichak, Edward A

    2018-02-01

    Clinical coding systems have been developed to translate real-world healthcare information such as prescriptions, diagnoses and procedures into standardized codes appropriate for use in large healthcare datasets. Due to the lack of information on coding system characteristics and insufficient uniformity in coding practices, there is a growing need for better understanding of coding systems and their use in pharmacoepidemiology and observational real world data research. To determine: 1) the number of available coding systems and their characteristics, 2) which pharmacoepidemiology databases are they adopted in, 3) what outcomes and exposures can be identified from each coding system, and 4) how robust they are with respect to consistency and validity in pharmacoepidemiology and observational database studies. Electronic literature database and unpublished literature searches, as well as hand searching of relevant journals were conducted to identify eligible articles discussing characteristics and applications of coding systems in use and published in the English language between 1986 and 2016. Characteristics considered included type of information captured by codes, clinical setting(s) of use, adoption by a pharmacoepidemiology database, region, and available mappings. Applications articles describing the use and validity of specific codes, code lists, or algorithms were also included. Data extraction was performed independently by two reviewers and a narrative synthesis was performed. A total of 897 unique articles and 57 coding systems were identified, 17% of which included country-specific modifications or multiple versions. Procedures (55%), diagnoses (36%), drugs (38%), and site of disease (39%) were most commonly and directly captured by these coding systems. The systems were used to capture information from the following clinical settings: inpatient (63%), ambulatory (55%), emergency department (ED, 34%), and pharmacy (13%). More than half of all coding

  4. The Code of Criminal Procedure Brazilian Military: Ode or Requiem for the Inquisitorial System?

    Directory of Open Access Journals (Sweden)

    Nicanor Henrique Netto Armando

    2015-12-01

    Full Text Available This paper conducted a literature review aimed to identify the epistemological principles, social and political underlying the inquisitorial and adversarial procedural systems, which connect the characteristics of each of these systems, to investigate which of them matches the Criminal Procedure Code of Military (CPPM. The findings of the survey show that for a system to be characterize as libelous not just the separation of functions to accuse and judge. It is necessary to keep the separation for which the structure of dialectical process does not break, and the evidential initiative is always reserved for the parties. With regard to systems,the sticking point is the identification of its principle informer, for it is he who will determine if the system is inquisition regime or libelous, and not ancillary elements (orality, advertising, separation  of  activities.  The  process  aims  to  reconstruction  of  a  historical  fact,  the management of proof is erected the core founding or the unifying principle of a system. In this context, devices in the CPPM to give judge powers investigative denote the adoption of the principle inquisitive, who founded a system inquisition regime, because they represent a breach of equality, contradictory, the very structure of the dialectic process. Consequently, fulminam the main guarantee of urisdiction, which is the impartiality of the judge. With this, it is concluded that the CPPM is essentially inquisition regime. This study also (rethinking the need of dichotomization of criminal proceedings Brazilian, perquirindo if the duality of legislation on criminal proceedings is consistent with a current view of the process as a constitutional  model  that  ensures  the  enforcement  of fundamental rights by bases principiológicas jointly agreed to any process.

  5. Consequences of the release of chemical pollutants on the transfers of radiioactive products in aquatic systems

    International Nuclear Information System (INIS)

    Bittel, R.

    1975-01-01

    With the increasing rate of industrial activities, aquatic systems undergo, more and more frequently, the accumulation of chemical and radioactive wastes released separatly or associated in the same discharge. An attempt is made to evaluate the consequence of the association of pollutants on the transfers of neutron activation radionuclides. Emphasis is given to heavy metal pollution and complexing agents [fr

  6. Environmental decision support system on base of geoinformational technologies for the analysis of nuclear accident consequences

    International Nuclear Information System (INIS)

    Haas, T.C.; Maigan, M.; Arutyunyan, R.V.; Bolshov, L.A.; Demianov, V.V.

    1996-01-01

    The report deals with description of the concept and prototype of environmental decision support system (EDSS) for the analysis of late off-site consequences of severe nuclear accidents and analysis, processing and presentation of spatially distributed radioecological data. General description of the available software, use of modem achievements of geostatistics and stochastic simulations for the analysis of spatial data are presented and discussed

  7. A comparison of thermal algorithms of fuel rod performance code systems

    International Nuclear Information System (INIS)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

  8. A comparison of thermal algorithms of fuel rod performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

  9. THIDA: code system for calculation of the exposure dose rate around a fusion device

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Igarashi, Masahito.

    1978-12-01

    A code system THIDA has been developed for calculation of the exposure dose rates around a fusion device. It consists of the following: one- and two-dimensional discrete ordinate transport codes; induced activity calculation code; activation chain, activation cross section, radionuclide gamma-ray energy/intensity and gamma-ray group constant files; and gamma ray flux to exposure dose rate conversion coefficients. (author)

  10. A code system to generate multigroup cross-sections using basic data

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

    For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

  11. Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Chong, E-mail: ch.zhou@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany); Yang Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Cheng Xu [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Modification of system code ATHLET for supercritical water application. Black-Right-Pointing-Pointer Development and assessment of a heat transfer package for supercritical water. Black-Right-Pointing-Pointer Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. Black-Right-Pointing-Pointer Application of the modified code to LOCA analysis of a supercritical water cooled in-pile fuel qualification test loop. - Abstract: Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards-O'Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the

  12. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  13. “PROCESS”: A systems code for fusion power plants—Part 1: Physics

    Energy Technology Data Exchange (ETDEWEB)

    Kovari, M., E-mail: michael.kovari@ccfe.ac.uk; Kemp, R.; Lux, H.; Knight, P.; Morris, J.; Ward, D.J.

    2014-12-15

    Highlights: • PROCESS is a fusion reactor systems code. • It optimises a figure of merit subject to constraints chosen by the user. • CCFE are working to make the assumptions and equations explicit and public. • The PROCESS homepage is (www.ccfe.ac.uk/powerplants.aspx). - Abstract: PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system, from the basic plasma physics to the generation of electricity. It has been used for many years, but details of its operation have not been previously published. This paper describes some of its capabilities. PROCESS is usually used in optimisation mode, in which it finds a set of parameters that maximise (or minimise) a figure of merit chosen by the user, while being consistent with the inputs and the specified constraints. Because the user can apply all the physically relevant constraints, while allowing a large number of parameters to vary, it is in principle only necessary to run the code once to produce a self-consistent, physically plausible reactor model. The scope of PROCESS is very wide and goes well beyond reactor physics, including conversion of heat to electricity, buildings, and costs, but this paper describes only the plasma physics and magnetic field calculations. The capabilities of PROCESS in plasma physics are limited, as its main aim is to combine engineering, physics and economics. A model is described which shows the main plasma features of an inductive ITER scenario. Significant differences between the PROCESS results and the published scenario include the bootstrap current and loop voltage. The PROCESS models for these are being revised. Two new models for DEMO have been obtained. The first, DEMO A, is intended to be “conservative” in that it might be possible to build it using the technology of the near future. For example, since current drive technologies are not yet

  14. Assessment of Consequences of Replacement of System of the Uniform Tax on Imputed Income Patent System of the Taxation

    Directory of Open Access Journals (Sweden)

    Galina A. Manokhina

    2012-11-01

    Full Text Available The article highlights the main questions concerning possible consequences of replacement of nowadays operating system in the form of a single tax in reference to imputed income with patent system of the taxation. The main advantages and drawbacks of new system of the taxation are shown, including the opinion that not the replacement of one special mode of the taxation with another is more effective, but the introduction of patent a taxation system as an auxilary system.

  15. ["Handle with care": about the potential unintended consequences of oracular artificial intelligence systems in medicine.

    Science.gov (United States)

    Cabitza, Federico; Alderighi, Camilla; Rasoini, Raffaele; Gensini, Gian Franco

    2017-10-01

    Decisional support systems based on machine learning (ML) in medicine are gaining a growing interest as some recent articles have highlighted the high diagnostic accuracy exhibited by these systems in specific medical contexts. However, it is implausible that any potential advantage can be obtained without some potential drawbacks. In light of the current gaps in medical research about the side effects of the application of these new AI systems in medical practice, in this article we summarize the main unexpected consequences that may result from the widespread application of "oracular" systems, that is highly accurate systems that cannot give reasonable explanations of their advice as those endowed with predictive models developed with ML techniques usually are. These consequences range from the intrinsic uncertainty in the data that are used to train and feed these systems, to the inadequate explainability of their output; through the risk of overreliance, deskilling and context desensitization of their end-users. Although some of these issues may be currently hard to evaluate due to the still scarce adoption of these decisional systems in medical practice, we advocate the study of these potential consequences also for a more informed policy of approval beyond hype and disenchantment.

  16. Coupled CFD - system-code simulation of a gas cooled reactor

    International Nuclear Information System (INIS)

    Yan, Yizhou; Rizwan-uddin

    2011-01-01

    A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)

  17. Near-Capacity Coding for Discrete Multitone Systems with Impulse Noise

    Directory of Open Access Journals (Sweden)

    Kschischang Frank R

    2006-01-01

    Full Text Available We consider the design of near-capacity-achieving error-correcting codes for a discrete multitone (DMT system in the presence of both additive white Gaussian noise and impulse noise. Impulse noise is one of the main channel impairments for digital subscriber lines (DSL. One way to combat impulse noise is to detect the presence of the impulses and to declare an erasure when an impulse occurs. In this paper, we propose a coding system based on low-density parity-check (LDPC codes and bit-interleaved coded modulation that is capable of taking advantage of the knowledge of erasures. We show that by carefully choosing the degree distribution of an irregular LDPC code, both the additive noise and the erasures can be handled by a single code, thus eliminating the need for an outer code. Such a system can perform close to the capacity of the channel and for the same redundancy is significantly more immune to the impulse noise than existing methods based on an outer Reed-Solomon (RS code. The proposed method has a lower implementation complexity than the concatenated coding approach.

  18. COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

    Directory of Open Access Journals (Sweden)

    Filip Osusky

    2018-05-01

    Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

  19. PRELIMINARY STUDY ON APPLICATION OF MAX PLUS ALGEBRA IN DISTRIBUTED STORAGE SYSTEM THROUGH NETWORK CODING

    Directory of Open Access Journals (Sweden)

    Agus Maman Abadi

    2016-04-01

    Full Text Available The increasing need in techniques of storing big data presents a new challenge. One way to address this challenge is the use of distributed storage systems. One strategy that implemented in distributed data storage systems is the use of Erasure Code which applied to network coding. The code used in this technique is based on the algebraic structure which is called as vector space. Some studies have also been carried out to create code that is based on other algebraic structures such as module.  In this study, we are going to try to set up a code based on the algebraic structure which is a generalization of the module that is semimodule by utilizing the max operations and sum operations at max plus algebra. The results of this study indicate that the max operation and the addition operation on max plus algebra cannot be used to establish a semimodule code, but by modifying the operation "+" as "min", we get a code based on semimodule. Keywords:   code, distributed storage systems, network coding, semimodule, max plus algebra

  20. 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.

  1. Flammable gas deflagration consequence calculations for the tank waste remediation system basis for interim operation

    Energy Technology Data Exchange (ETDEWEB)

    Van Vleet, R.J., Westinghouse Hanford

    1996-08-13

    This paper calculates the radiological dose consequences and the toxic exposures for deflagration accidents at various Tank Waste Remediation System facilities. These will be used in support of the Tank Waste Remediation System Basis for Interim Operation.The attached SD documents the originator`s analysis only. It shall not be used as the final or sole document for effecting changes to an authorization basis or safety basis for a facility or activity.

  2. APPC - A new standardised coding system for trans-organisational PACS retrieval

    International Nuclear Information System (INIS)

    Fruehwald, F.; Lindner, A.; Mostbeck, G.; Hruby, W.; Fruehwald-Pallamar, J.

    2010-01-01

    As part of a general strategy to integrate the health care enterprise, Austria plans to connect the Picture Archiving and Communication Systems (PACS) of all radiological institutions into a nationwide network. To facilitate the search for relevant correlative imaging data in the PACS of different organisations, a coding system was compiled for all radiological procedures and necessary anatomical details. This code, called the Austrian PACS Procedure Code (APPC), was granted the status of a standard under HL7. Examples are provided of effective coding and filtering when searching for relevant imaging material using the APPC, as well as the planned process for future adjustments of the APPC. The implementation and how the APPC will fit into the future electronic environment, which will include an electronic health act for all citizens in Austria, are discussed. A comparison to other nationwide electronic health record projects and coding systems is given. Limitations and possible use in physical storage media are contemplated. (orig.)

  3. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Y. M.; Park, H. S.; Kim, T. H.

    2015-01-01

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software

  4. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. M.; Park, H. S. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

    2015-10-15

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software.

  5. Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

  6. Implantation of Magint-Maggraf code in the Microvax-3600 system

    International Nuclear Information System (INIS)

    Leite Neto, J.P.; Shiabata, C.S.; Montes, A.

    1990-01-01

    An auxiliary code, named MAGGRAF, was developed and implemented on the MicroVAX-3600 system of the Plasma Laboratory at the Institute for Space Research, in order to perform the graphical output of the numerical results provided by the magnetic field calculation code MAGINT. In this report we present a brief description of the graphical code, of the parameters which specify the different output options, and of the structure of the data file containing these parameters. Some examples are shown to illustrate the versatility of the code, as well as the quality of the graphs. (author)

  7. Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

    Science.gov (United States)

    Matem, Rima; Aljunid, S. A.; Junita, M. N.; Rashidi, C. B. M.; Shihab Aqrab, Israa

    2017-11-01

    In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN), shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.

  8. Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

    Directory of Open Access Journals (Sweden)

    Matem Rima

    2017-01-01

    Full Text Available In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN, shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.

  9. Optimization of multi-phase compressible lattice Boltzmann codes on massively parallel multi-core systems

    NARCIS (Netherlands)

    Biferale, L.; Mantovani, F.; Pivanti, M.; Pozzati, F.; Sbragaglia, M.; Schifano, S.F.; Toschi, F.; Tripiccione, R.

    2011-01-01

    We develop a Lattice Boltzmann code for computational fluid-dynamics and optimize it for massively parallel systems based on multi-core processors. Our code describes 2D multi-phase compressible flows. We analyze the performance bottlenecks that we find as we gradually expose a larger fraction of

  10. Applications of the lots computer code to laser fusion systems and other physical optics problems

    International Nuclear Information System (INIS)

    Lawrence, G.; Wolfe, P.N.

    1979-01-01

    The Laser Optical Train Simulation (LOTS) code has been developed at the Optical Sciences Center, University of Arizona under contract to Los Alamos Scientific Laboratory (LASL). LOTS is a diffraction based code designed to beam quality and energy of the laser fusion system in an end-to-end calculation

  11. Information rates of next-generation long-haul optical fiber systems using coded modulation

    NARCIS (Netherlands)

    Liga, G.; Alvarado, A.; Agrell, E.; Bayvel, P.

    2017-01-01

    A comprehensive study of the coded performance of long-haul spectrally-efficient WDM optical fiber transmission systems with different coded modulation decoding structures is presented. Achievable information rates are derived for three different square QAM formats and the optimal format is

  12. A QR code identification technology in package auto-sorting system

    Science.gov (United States)

    di, Yi-Juan; Shi, Jian-Ping; Mao, Guo-Yong

    2017-07-01

    Traditional manual sorting operation is not suitable for the development of Chinese logistics. For better sorting packages, a QR code recognition technology is proposed to identify the QR code label on the packages in package auto-sorting system. The experimental results compared with other algorithms in literatures demonstrate that the proposed method is valid and its performance is superior to other algorithms.

  13. Integration of QR codes into an anesthesia information management system for resident case log management.

    Science.gov (United States)

    Avidan, Alexander; Weissman, Charles; Levin, Phillip D

    2015-04-01

    Quick response (QR) codes containing anesthesia syllabus data were introduced into an anesthesia information management system. The code was generated automatically at the conclusion of each case and available for resident case logging using a smartphone or tablet. The goal of this study was to evaluate the use and usability/user-friendliness of such system. Resident case logging practices were assessed prior to introducing the QR codes. QR code use and satisfactions amongst residents was reassessed at three and six months. Before QR code introduction only 12/23 (52.2%) residents maintained a case log. Most of the remaining residents (9/23, 39.1%) expected to receive a case list from the anesthesia information management system database at the end of their residency. At three months and six months 17/26 (65.4%) and 15/25 (60.0%) residents, respectively, were using the QR codes. Satisfaction was rated as very good or good. QR codes for residents' case logging with smartphones or tablets were successfully introduced in an anesthesia information management system and used by most residents. QR codes can be successfully implemented into medical practice to support data transfer. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  14. 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.

  15. Demonstration of Vibrational Braille Code Display Using Large Displacement Micro-Electro-Mechanical Systems Actuators

    Science.gov (United States)

    Watanabe, Junpei; Ishikawa, Hiroaki; Arouette, Xavier; Matsumoto, Yasuaki; Miki, Norihisa

    2012-06-01

    In this paper, we present a vibrational Braille code display with large-displacement micro-electro-mechanical systems (MEMS) actuator arrays. Tactile receptors are more sensitive to vibrational stimuli than to static ones. Therefore, when each cell of the Braille code vibrates at optimal frequencies, subjects can recognize the codes more efficiently. We fabricated a vibrational Braille code display that used actuators consisting of piezoelectric actuators and a hydraulic displacement amplification mechanism (HDAM) as cells. The HDAM that encapsulated incompressible liquids in microchambers with two flexible polymer membranes could amplify the displacement of the MEMS actuator. We investigated the voltage required for subjects to recognize Braille codes when each cell, i.e., the large-displacement MEMS actuator, vibrated at various frequencies. Lower voltages were required at vibration frequencies higher than 50 Hz than at vibration frequencies lower than 50 Hz, which verified that the proposed vibrational Braille code display is efficient by successfully exploiting the characteristics of human tactile receptors.

  16. 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...

  17. A Mechanism to Avoid Collusion Attacks Based on Code Passing in Mobile Agent Systems

    Science.gov (United States)

    Jaimez, Marc; Esparza, Oscar; Muñoz, Jose L.; Alins-Delgado, Juan J.; Mata-Díaz, Jorge

    Mobile agents are software entities consisting of code, data, state and itinerary that can migrate autonomously from host to host executing their code. Despite its benefits, security issues strongly restrict the use of code mobility. The protection of mobile agents against the attacks of malicious hosts is considered the most difficult security problem to solve in mobile agent systems. In particular, collusion attacks have been barely studied in the literature. This paper presents a mechanism that avoids collusion attacks based on code passing. Our proposal is based on a Multi-Code agent, which contains a different variant of the code for each host. A Trusted Third Party is responsible for providing the information to extract its own variant to the hosts, and for taking trusted timestamps that will be used to verify time coherence.

  18. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    Science.gov (United States)

    Karim, Julia Abdul

    2008-05-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

  19. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    International Nuclear Information System (INIS)

    Karim, Julia Abdul

    2008-01-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained

  20. A good performance watermarking LDPC code used in high-speed optical fiber communication system

    Science.gov (United States)

    Zhang, Wenbo; Li, Chao; Zhang, Xiaoguang; Xi, Lixia; Tang, Xianfeng; He, Wenxue

    2015-07-01

    A watermarking LDPC code, which is a strategy designed to improve the performance of the traditional LDPC code, was introduced. By inserting some pre-defined watermarking bits into original LDPC code, we can obtain a more correct estimation about the noise level in the fiber channel. Then we use them to modify the probability distribution function (PDF) used in the initial process of belief propagation (BP) decoding algorithm. This algorithm was tested in a 128 Gb/s PDM-DQPSK optical communication system and results showed that the watermarking LDPC code had a better tolerances to polarization mode dispersion (PMD) and nonlinearity than that of traditional LDPC code. Also, by losing about 2.4% of redundancy for watermarking bits, the decoding efficiency of the watermarking LDPC code is about twice of the traditional one.

  1. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  2. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  3. The symbol coding language for the BUTs processor of in-core reactor control systems

    International Nuclear Information System (INIS)

    Vorob'ev, D.M.; Golovanov, M.N.; Levin, G.L.; Parfenova, T.K.; Filatov, V.P.

    1978-01-01

    A symbolic coding language is described; it has been developed for automation of making up programs for in-core control systems. The systems use the ideology of the CAMAC-VECTOR system and include the BUTs-20 processor. The symbolic coding language has been developed as a programming language of the ASSEMBLER type. Operators of instructions and pseudo-instructions, the rules of reading in the text of the source program, and operator record formats are considered

  4. Use the Bar Code System to Improve Accuracy of the Patient and Sample Identification.

    Science.gov (United States)

    Chuang, Shu-Hsia; Yeh, Huy-Pzu; Chi, Kun-Hung; Ku, Hsueh-Chen

    2018-01-01

    In time and correct sample collection were highly related to patient's safety. The sample error rate was 11.1%, because misbranded patient information and wrong sample containers during January to April, 2016. We developed a barcode system of "Specimens Identify System" through process of reengineering of TRM, used bar code scanners, add sample container instructions, and mobile APP. Conclusion, the bar code systems improved the patient safety and created green environment.

  5. CORESAFE: A Formal Approach against Code Replacement Attacks on Cyber Physical Systems

    Science.gov (United States)

    2018-04-19

    AFRL-AFOSR-JP-TR-2018-0035 CORESAFE:A Formal Approach against Code Replacement Attacks on Cyber Physical Systems Sandeep Shukla INDIAN INSTITUTE OF...Formal Approach against Code Replacement Attacks on Cyber Physical Systems 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA2386-16-1-4099 5c.  PROGRAM ELEMENT...SUPPLEMENTARY NOTES 14.  ABSTRACT Industrial Control Systems (ICS) used in manufacturing, power generators and other critical infrastructure monitoring and

  6. Development of a system of computer codes for severe accident analyses and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    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. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

  7. Feasibility study on the rod ejection accident analysis with RETRAN-MASTER code system

    International Nuclear Information System (INIS)

    Kim, Y. H.; Lee, C. S.

    2003-01-01

    KEPRI has been developed the in-house methodology for non-LOCA safety analyses based on the codes and methodologies of vendors and EPRI. Using the methodology, the rod ejection accident, which is classified into the generic accident analysis category of reactivity insertion accident in primary system, has been analyzed with RETRAN-MASTER code system. And the feasibility of the coupled code system has been verified by the review of the results. Furthermore, to assess the important parameters to the accident, the sensitivity analyses have been carried out over some parameters

  8. Development of a system of computer codes for severe accident analyses and its applications

    International Nuclear Information System (INIS)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

    1991-12-01

    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. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy

  9. Development of the computer code system for the analyses of PWR core

    International Nuclear Information System (INIS)

    Tsujimoto, Iwao; Naito, Yoshitaka.

    1992-11-01

    This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

  10. Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach

    International Nuclear Information System (INIS)

    Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou

    2005-01-01

    A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)

  11. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  12. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

    Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

  13. Addressing the scaling issue by thermalhydraulic system codes: recent results

    International Nuclear Information System (INIS)

    D'auria, F.; Cherubini, M.; Galassi, G.M.; Muellner, N. . E-mail of corresponding author: dauria@ing.unipi.it; D'auria, F.)

    2005-01-01

    This lecture presents an introduction into the scaling issue following a 'top-down' approach. This means, recent studies which deal with a scaling analysis in LWR with special regards to the WWER Russian reactor type are presented to demonstrate important phenomena for scaling, to be more precise, the counterpart test (CT) methodology. As an example, one CT, a Small Break LOCA carried out in the PSB facility, is presented. PSB is a full height, full pressure rig that reproduces a WWER 1000, power and volume scaling factor is 1:300. The CT has been designed deriving boundary and initial condition from the same test performed in LOBI (that reproduces a PWR). The adopted scaling approach is based on the selection of a few characteristic parameters. They are chosen taking into account their relevance in the behaviour of the transient. The calculation of the SBLOCA has been performed using Relap5/Mod3.3 computer code and its accuracy has been demonstrated by qualitative and quantitative evaluation. For the quantitative evaluation the use of the FFT Based Method is foreseen and the fulfilment of its limits has been obtained. The aim of the example is to give an overview about the theoretical concepts of scaling, which is termed the s caling strategy , and comprises the steps of the selected scaling approach. At the same time interesting results from ongoing research projects will be presented. Comparing experimental data it was found that the investigated facilities show similar behaviour concerning the time trends, and are able to predict on a qualitative level the same thermal hydraulic phenomena. Main obtained results are summarized as follows: PSB and LOBI main parameters have similar trends. This is a confirmation of the validity of the adopted scaling approach and shows that PWR and WWER reactor type behaviour are very close to each other. No new phenomena occur during the CT, notwithstanding the two facilities have a different lay out, and the already known

  14. Comparison of INTERA and WISAP consequence model application. Assessment of effectiveness of geologic isolation systems

    International Nuclear Information System (INIS)

    Cole, C.R.; Bond, F.W.

    1980-01-01

    The Waste Isolation Safety Assessment Program (WISAP) is being conducted to develop, for the Office of Nuclear Waste Isolation (ONWI), the methodology necessary to perform long-term safety assessments of deep geologic repositories. The Waste Isolation Pilot Plant (WIPP) program is developing a nuclear waste storage facility and is performing assessments of that site. WISAP and WIPP have similar, though independent, methodologies for assessing the consequences of a repository breach subsequent to closure. Intera Environmental Consultants are under contract to Sandia Laboratories to conduct the hydrologic and transport modeling for the WIPP Site Release Consequence Analysis (WIPP EIS/ER 1978). To provide a mutual benchmark check of the radionuclide and ground-water transport models of these two programs, ONWI has requested WISAP to perform a release consequence analysis based on the WIPP site, utilizing the same data and conceptual model which the WIPP program used for its environmental assessments. Therefore, only a portion of the WISAP methodology was used; specifically, only WISAP geotransport models were exercised. The other important parts of WISAP assessment methodology were not used, so that WISAP did not develop the scenario nor did WISAP interpret the field data to develop the conceptual model of the geohydrology of the WIPP site. The results of the comparative assessment are presented. Although the different models required slightly different input parameters, the results of the hydrologic simulations show a very close correspondence between the WISAP and WIPP predictions. This was as expected, since the various hydrologic codes available essentially utilize and solve the same basic flow equations. In addition, this report presents the results of the WISAP radionuclide transport model simulations. These results will provide the basis for comparison with WIPP results when these become available

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

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

    SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

  16. Investigation and proposal of the system to affect nuclear fuel type authorization and analysis code certification

    International Nuclear Information System (INIS)

    2006-03-01

    In order to develop the system to affect more advanced and rational regulations of nuclear fuels and earlier introduction of new technologies in nuclear power plants, domestic and overseas safety regulation systems and state of their implementation for water cooled reactor fuel and safety analysis code had been investigated and new regulation system to affect nuclear fuel type authorization and analysis code certification was proposed. Topical report system for common parts related with nuclear fuel type authorization and analysis code certification was firstly proposed for knowledge base. Maintaining consistent safety examination supported by experts, introduction of domestic efficient system for lead irradiation test fuel, and analysis code certification and quality assurance were also proposed. (T. Tanaka)

  17. Cardinality enhancement utilizing Sequential Algorithm (SeQ code in OCDMA system

    Directory of Open Access Journals (Sweden)

    Fazlina C. A. S.

    2017-01-01

    Full Text Available Optical Code Division Multiple Access (OCDMA has been important with increasing demand for high capacity and speed for communication in optical networks because of OCDMA technique high efficiency that can be achieved, hence fibre bandwidth is fully used. In this paper we will focus on Sequential Algorithm (SeQ code with AND detection technique using Optisystem design tool. The result revealed SeQ code capable to eliminate Multiple Access Interference (MAI and improve Bit Error Rate (BER, Phase Induced Intensity Noise (PIIN and orthogonally between users in the system. From the results, SeQ shows good performance of BER and capable to accommodate 190 numbers of simultaneous users contrast with existing code. Thus, SeQ code have enhanced the system about 36% and 111% of FCC and DCS code. In addition, SeQ have good BER performance 10-25 at 155 Mbps in comparison with 622 Mbps, 1 Gbps and 2 Gbps bit rate. From the plot graph, 155 Mbps bit rate is suitable enough speed for FTTH and LAN networks. Resolution can be made based on the superior performance of SeQ code. Thus, these codes will give an opportunity in OCDMA system for better quality of service in an optical access network for future generation's usage

  18. Design evaluation on sodium piping system and comparison of the design codes

    International Nuclear Information System (INIS)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

    2015-01-01

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  19. Design evaluation on sodium piping system and comparison of the design codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

    2015-03-15

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  20. Cardinality enhancement utilizing Sequential Algorithm (SeQ) code in OCDMA system

    Science.gov (United States)

    Fazlina, C. A. S.; Rashidi, C. B. M.; Rahman, A. K.; Aljunid, S. A.

    2017-11-01

    Optical Code Division Multiple Access (OCDMA) has been important with increasing demand for high capacity and speed for communication in optical networks because of OCDMA technique high efficiency that can be achieved, hence fibre bandwidth is fully used. In this paper we will focus on Sequential Algorithm (SeQ) code with AND detection technique using Optisystem design tool. The result revealed SeQ code capable to eliminate Multiple Access Interference (MAI) and improve Bit Error Rate (BER), Phase Induced Intensity Noise (PIIN) and orthogonally between users in the system. From the results, SeQ shows good performance of BER and capable to accommodate 190 numbers of simultaneous users contrast with existing code. Thus, SeQ code have enhanced the system about 36% and 111% of FCC and DCS code. In addition, SeQ have good BER performance 10-25 at 155 Mbps in comparison with 622 Mbps, 1 Gbps and 2 Gbps bit rate. From the plot graph, 155 Mbps bit rate is suitable enough speed for FTTH and LAN networks. Resolution can be made based on the superior performance of SeQ code. Thus, these codes will give an opportunity in OCDMA system for better quality of service in an optical access network for future generation's usage

  1. Design of Spreading-Codes-Assisted Active Imaging System

    Directory of Open Access Journals (Sweden)

    Alexey Volkov

    2015-07-01

    Full Text Available This work discusses an innovative approach to imaging which can improve the robustness of existing active-range measurement methods and potentially enhance their use in a variety of outdoor applications. By merging a proven modulation technique from the domain of spread-spectrum communications with the bleeding-edge CMOS sensor technology, the prototype of the modulated range sensor is designed and evaluated. A suitable set of application-specific spreading codes is proposed, evaluated and tested on the prototype. Experimental results show that the introduced modulation technique significantly reduces the impacts of environmental factors such as sunlight and external light sources, as well as mutual interference of identical devices. The proposed approach can be considered as a promising basis for a new generation of robust and cost-efficient range-sensing solutions for automotive applications, autonomous vehicles or robots.

  2. Improving performance of DS-CDMA systems using chaotic complex Bernoulli spreading codes

    Science.gov (United States)

    Farzan Sabahi, Mohammad; Dehghanfard, Ali

    2014-12-01

    The most important goal of spreading spectrum communication system is to protect communication signals against interference and exploitation of information by unintended listeners. In fact, low probability of detection and low probability of intercept are two important parameters to increase the performance of the system. In Direct Sequence Code Division Multiple Access (DS-CDMA) systems, these properties are achieved by multiplying the data information in spreading sequences. Chaotic sequences, with their particular properties, have numerous applications in constructing spreading codes. Using one-dimensional Bernoulli chaotic sequence as spreading code is proposed in literature previously. The main feature of this sequence is its negative auto-correlation at lag of 1, which with proper design, leads to increase in efficiency of the communication system based on these codes. On the other hand, employing the complex chaotic sequences as spreading sequence also has been discussed in several papers. In this paper, use of two-dimensional Bernoulli chaotic sequences is proposed as spreading codes. The performance of a multi-user synchronous and asynchronous DS-CDMA system will be evaluated by applying these sequences under Additive White Gaussian Noise (AWGN) and fading channel. Simulation results indicate improvement of the performance in comparison with conventional spreading codes like Gold codes as well as similar complex chaotic spreading sequences. Similar to one-dimensional Bernoulli chaotic sequences, the proposed sequences also have negative auto-correlation. Besides, construction of complex sequences with lower average cross-correlation is possible with the proposed method.

  3. 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.

  4. Modelling of the Rod Control System in the coupled code RELAP5-QUABOX/CUBBOX

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Grgic, D.

    1999-01-01

    There is a general agreement that for many light water reactor transient calculations, it is necessary to use a multidimensional neutron kinetics model coupled to sophisticated thermal-hydraulic models in order to obtain satisfactory results. These calculations are needed for a variety of applications for licensing safety analyses, probabilistic risk assessment, operational support, and training. At FER, Zagreb, a coupling of 3D neutronics code QUABOX/CUBBOX and system code RELAP5 was performed. In the paper the Rod Control System model in the RELAP5 part of the coupled code is presented. A first testing of the model was performed by calculation of reactor trip from full power for NPP Krsko. Results of 3D neutronics calculation obtained by coupled code QUABOX/CUBBOX were compared with point kinetics calculation performed with RELAP5 stand alone code.(author)

  5. Simulations of X-ray synchrotron beams using the EGS4 code system in medical applications

    International Nuclear Information System (INIS)

    Orion, I.; Henn, A.; Sagi, I.; Dilmanian, F.A.; Pena, L.; Rosenfeld, A.B.

    2001-01-01

    X-ray synchrotron beams are commonly used in biological and medical research. The availability of intense, polarized low-energy photons from the synchrotron beams provides a high dose transfer to biological materials. The EGS4 code system, which includes the photoelectron angular distribution, electron motion inside a magnetic field, and the LSCAT package, found to be the appropriate Monte Carlo code for synchrotron-produced X-ray simulations. The LSCAT package was developed in 1995 for the EGS4 code to contain the routines to simulate the linear polarization, the bound Compton, and the incoherent scattering functions. Three medical applications were demonstrated using the EGS4 Monte Carlo code as a proficient simulation code system for the synchrotron low-energy X-ray source. (orig.)

  6. Systems engineering applied to integrated safety management for high consequence facilities

    International Nuclear Information System (INIS)

    Barter, R; Morais, B.

    1998-01-01

    Integrated Safety Management is a concept that is being actively promoted by the U.S. Department of Energy as a means of assuring safe operation of its facilities. The concept involves the integration of safety precepts into work planning rather than adjusting for safe operations after defining the work activity. The system engineering techniques used to design an integrated safety management system for a high consequence research facility are described. An example is given to show how the concepts evolved with the system design

  7. Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

    International Nuclear Information System (INIS)

    Risken, Tobias; Koch, Marco K.

    2011-01-01

    The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

  8. PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

    International Nuclear Information System (INIS)

    Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

    1995-01-01

    1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

  9. Performance of asynchronous fiber-optic code division multiple access system based on three-dimensional wavelength/time/space codes and its link analysis.

    Science.gov (United States)

    Singh, Jaswinder

    2010-03-10

    A novel family of three-dimensional (3-D) wavelength/time/space codes for asynchronous optical code-division-multiple-access (CDMA) systems with "zero" off-peak autocorrelation and "unity" cross correlation is reported. Antipodal signaling and differential detection is employed in the system. A maximum of [(W x T+1) x W] codes are generated for unity cross correlation, where W and T are the number of wavelengths and time chips used in the code and are prime. The conditions for violation of the cross-correlation constraint are discussed. The expressions for number of generated codes are determined for various code dimensions. It is found that the maximum number of codes are generated for S systems. The codes have a code-set-size to code-size ratio greater than W/S. For instance, with a code size of 2065 (59 x 7 x 5), a total of 12,213 users can be supported, and 130 simultaneous users at a bit-error rate (BER) of 10(-9). An arrayed-waveguide-grating-based reconfigurable encoder/decoder design for 2-D implementation for the 3-D codes is presented so that the need for multiple star couplers and fiber ribbons is eliminated. The hardware requirements of the coders used for various modulation/detection schemes are given. The effect of insertion loss in the coders is shown to be significantly reduced with loss compensation by using an amplifier after encoding. An optical CDMA system for four users is simulated and the results presented show the improvement in performance with the use of loss compensation.

  10. The Conversion of Wiswesser Line Notations to Ring Codes. I. The Conversion of Ring Systems

    Science.gov (United States)

    Granito, Charles E.; And Others

    1972-01-01

    The computerized conversion of Wiswesser Line Notations to Ring Codes, using a two-part approach, and the set of computer programs generated for the conversion of ring systems are described. (9 references) (Author)

  11. A systematic literature review of automated clinical coding and classification systems.

    Science.gov (United States)

    Stanfill, Mary H; Williams, Margaret; Fenton, Susan H; Jenders, Robert A; Hersh, William R

    2010-01-01

    Clinical coding and classification processes transform natural language descriptions in clinical text into data that can subsequently be used for clinical care, research, and other purposes. This systematic literature review examined studies that evaluated all types of automated coding and classification systems to determine the performance of such systems. Studies indexed in Medline or other relevant databases prior to March 2009 were considered. The 113 studies included in this review show that automated tools exist for a variety of coding and classification purposes, focus on various healthcare specialties, and handle a wide variety of clinical document types. Automated coding and classification systems themselves are not generalizable, nor are the results of the studies evaluating them. Published research shows these systems hold promise, but these data must be considered in context, with performance relative to the complexity of the task and the desired outcome.

  12. Validation Study of CODES Dragonfly Network Model with Theta Cray XC System

    Energy Technology Data Exchange (ETDEWEB)

    Mubarak, Misbah [Argonne National Lab. (ANL), Argonne, IL (United States); Ross, Robert B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-05-31

    This technical report describes the experiments performed to validate the MPI performance measurements reported by the CODES dragonfly network simulation with the Theta Cray XC system at the Argonne Leadership Computing Facility (ALCF).

  13. An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

    International Nuclear Information System (INIS)

    Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

    2006-01-01

    KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

  14. Code division multiple-access techniques in optical fiber networks. II - Systems performance analysis

    Science.gov (United States)

    Salehi, Jawad A.; Brackett, Charles A.

    1989-08-01

    A technique based on optical orthogonal codes was presented by Salehi (1989) to establish a fiber-optic code-division multiple-access (FO-CDMA) communications system. The results are used to derive the bit error rate of the proposed FO-CDMA system as a function of data rate, code length, code weight, number of users, and receiver threshold. The performance characteristics for a variety of system parameters are discussed. A means of reducing the effective multiple-access interference signal by placing an optical hard-limiter at the front end of the desired optical correlator is presented. Performance calculations are shown for the FO-CDMA with an ideal optical hard-limiter, and it is shown that using a optical hard-limiter would, in general, improve system performance.

  15. Adaptive Multi-Layered Space-Time Block Coded Systems in Wireless Environments

    KAUST Repository

    Al-Ghadhban, Samir

    2014-01-01

    © 2014, Springer Science+Business Media New York. Multi-layered space-time block coded systems (MLSTBC) strike a balance between spatial multiplexing and transmit diversity. In this paper, we analyze the block error rate performance of MLSTBC

  16. Learning of spatio-temporal codes in a coupled oscillator system.

    Science.gov (United States)

    Orosz, Gábor; Ashwin, Peter; Townley, Stuart

    2009-07-01

    In this paper, we consider a learning strategy that allows one to transmit information between two coupled phase oscillator systems (called teaching and learning systems) via frequency adaptation. The dynamics of these systems can be modeled with reference to a number of partially synchronized cluster states and transitions between them. Forcing the teaching system by steady but spatially nonhomogeneous inputs produces cyclic sequences of transitions between the cluster states, that is, information about inputs is encoded via a "winnerless competition" process into spatio-temporal codes. The large variety of codes can be learned by the learning system that adapts its frequencies to those of the teaching system. We visualize the dynamics using "weighted order parameters (WOPs)" that are analogous to "local field potentials" in neural systems. Since spatio-temporal coding is a mechanism that appears in olfactory systems, the developed learning rules may help to extract information from these neural ensembles.

  17. Manual for IRS Coding. Joint IAEA/NEA International Reporting System for Operating Experience

    International Nuclear Information System (INIS)

    2011-01-01

    The International Reporting System for Operating Experience (IRS) is jointly operated by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA). In early 2010, the IAEA and OECD/NEA jointly issued the IRS Guidelines, which described the reporting system and process and gave users the necessary elements to enable them to produce IRS reports to a high standard of quality while retaining the effectiveness of the system expected by all Member States operating nuclear power plants. The purpose of the present Manual for IRS Coding is to provide supplementary guidance specifically on the coding element of IRS reports to ensure uniform coding of events that are reported through IRS. This Coding Manual does not supersede the IRS Guidelines, but rather, supports users and preparers in achieving a consistent and high level of quality in their IRS reports. Consistency and high quality in the IRS reports allow stakeholders to search and retrieve specific event information with ease. In addition, well-structured reports also enhance the efficient management of the IRS database. This Coding Manual will give specific guidance on the application of each section of the IRS codes, with examples where necessary, of when and how these codes are to be applied. As this reporting system is owned by the Member States, this manual has been developed and approved by the IRS National Coordinators with the assistance of the IAEA and NEA secretariats

  18. Code development of the national hemovigilance system and expansion strategies for hospital blood banks

    Directory of Open Access Journals (Sweden)

    Kim Jeongeun

    2012-01-01

    Full Text Available Objectives : The aims of this study were to develop reportable event codes that are applicable to the national hemovigilance systems for hospital blood banks, and to present expansion strategies for the blood banks. Materials and Methods : The data were obtained from a literature review and expert consultation, followed by adding to and revising the established hemovigilance code system and guidelines to develop reportable event codes for hospital blood banks. The Medical Error Reporting System-Transfusion Medicine developed in the US and other codes of reportable events were added to the Korean version of the Biologic Products Deviation Report (BPDR developed by the Korean Red Cross Blood Safety Administration, then using these codes, mapping work was conducted. We deduced outcomes suitable for practice, referred to the results of the advisory councils, and conducted a survey with experts and blood banks practitioners. Results : We developed reportable event codes that were applicable to hospital blood banks and could cover blood safety - from blood product safety to blood transfusion safety - and also presented expansion strategies for hospital blood banks. Conclusion : It was necessary to add 10 major categories to the blood transfusion safety stage and 97 reportable event codes to the blood safety stage. Contextualized solutions were presented on 9 categories of expansion strategies of hemovigilance system for the hospital blood banks.

  19. Development of the next generation code system as an engineering modeling language. (2). Study with prototyping

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto

    2003-04-01

    In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomena to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical properties and engineering models in many different fields. Aiming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1) Multi-language (SoftWIRE.NET, Visual Basic.NET and Fortran) (2) Fortran 90 and (3) Python to make a prototype of the next generation code system. As this result, the followings were confirmed. (1) It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using Visual Basic.NET. (2) The maintainability of the existing code written by Fortran 77 can be improved by using the new features of Fortran 90. (3) The toolbox-type code system can be built by using Python. (author)

  20. 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.

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

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

    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

  2. Health systems as defences against the consequences of poverty: equity in health as social justice.

    Science.gov (United States)

    Mburu, F M

    1983-01-01

    The main development problems in the Third World are known to be gross socioeconomic inequality, widespread poor health status accompanied by high fertility and infant mortality rates, low life expectancy, mass illiteracy and mass poverty. In most of these countries governments invest a great deal of scarce resources toward the consequences of poverty rather than it causes. The paucity of resources for such social services is exacerbated by continuously increasing demands and needs which have to be satisfied. Unmet needs tend to cause apathy in the population. For purposes of controlling poverty and its consequences, these must be clearly formulated and relevant policies, a commitment to implement such policies, adequate administrative capacity and reasonably adequate resources. In the case of the health services system, the same requirements apply. Above all, the health system has to be directed toward the greatest needs of the population. This must involve policy makers, implementors and the consumer community. This paper argues that health systems cannot be an effective weapon against the consequences of poverty unless the above kinds of policy exist and are implemented.

  3. Use of source term code package in the ELEBRA MX-850 system

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.; Goes, A.G.A.

    1988-12-01

    The implantation of source term code package in the ELEBRA-MX850 system is presented. The source term is formed when radioactive materials generated in nuclear fuel leakage toward containment and the external environment to reactor containment. The implantated version in the ELEBRA system are composed of five codes: MARCH 3, TRAPMELT 3, THCCA, VANESA and NAVA. The original example case was used. The example consists of a small loca accident in a PWR type reactor. A sensitivity study for the TRAPMELT 3 code was carried out, modifying the 'TIME STEP' to estimate the processing time of CPU for executing the original example case. (M.C.K.) [pt

  4. Integrated transport code system for a multicomponent plasma in a gas dynamic trap

    International Nuclear Information System (INIS)

    Anikeev, A.V.; Karpushov, A.N.; Noak, K.; Strogalova, S.L.

    2000-01-01

    This report is focused on the development of the theoretical and numerical models of multicomponent high-β plasma confinement and transport in the gas-dynamic trap (GDT). In order to simulate the plasma behavior in the GDT as well as that in the GDT-based neutron source the Integrated Transport Code System is developed from existing stand-alone codes calculating the target plasma, the fast ions and the neutral gas in the GDT. The code system considers the full dependence of the transport phenomena on space, time, energy and angle variables as well as the interactions between the particle fields [ru

  5. PROMETHEUS - a code system for dynamic 3-D analysis of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khotylev, V.A.; Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-09-01

    The paper presents a multidimensional, general-purpose neutronics code system. It solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three-dimensional geometry. Due to a number of specialized features such as cavity treatment, automated convergence control, burnup treatment using the full isotopic transition matrix, the code system can be applied for the analysis of fast and slow transients in small, large, and innovative reactor cores. (author)

  6. Power Optimization of Wireless Media Systems With Space-Time Block Codes

    OpenAIRE

    Yousefi'zadeh, Homayoun; Jafarkhani, Hamid; Moshfeghi, Mehran

    2004-01-01

    We present analytical and numerical solutions to the problem of power control in wireless media systems with multiple antennas. We formulate a set of optimization problems aimed at minimizing total power consumption of wireless media systems subject to a given level of QoS and an available bit rate. Our formulation takes in to consideration the power consumption related to source coding, channel coding, and transmission of multiple-transmit antennas. In our study, we consider Gauss-Markov and...

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

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    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

  8. Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

    International Nuclear Information System (INIS)

    Kirk, B.L.; West, J.T.

    1984-06-01

    The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

  9. Dimensioning BCH codes for coherent DQPSK systems with laser phase noise and cycle slips

    DEFF Research Database (Denmark)

    Leong, Miu Yoong; Larsen, Knud J.; Jacobsen, Gunnar

    2014-01-01

    Forward error correction (FEC) plays a vital role in coherent optical systems employing multi-level modulation. However, much of coding theory assumes that additive white Gaussian noise (AWGN) is dominant, whereas coherent optical systems have significant phase noise (PN) in addition to AWGN...... approach for a target post-FEC BER of 10-5. Codes dimensioned with our bivariate binomial model meet the target within 0.2-dB signal-to-noise ratio....

  10. History by history statistical estimators in the BEAM code system

    International Nuclear Information System (INIS)

    Walters, B.R.B.; Kawrakow, I.; Rogers, D.W.O.

    2002-01-01

    A history by history method for estimating uncertainties has been implemented in the BEAMnrc and DOSXYZnrc codes replacing the method of statistical batches. This method groups scored quantities (e.g., dose) by primary history. When phase-space sources are used, this method groups incident particles according to the primary histories that generated them. This necessitated adding markers (negative energy) to phase-space files to indicate the first particle generated by a new primary history. The new method greatly reduces the uncertainty in the uncertainty estimate. The new method eliminates one dimension (which kept the results for each batch) from all scoring arrays, resulting in memory requirement being decreased by a factor of 2. Correlations between particles in phase-space sources are taken into account. The only correlations with any significant impact on uncertainty are those introduced by particle recycling. Failure to account for these correlations can result in a significant underestimate of the uncertainty. The previous method of accounting for correlations due to recycling by placing all recycled particles in the same batch did work. Neither the new method nor the batch method take into account correlations between incident particles when a phase-space source is restarted so one must avoid restarts

  11. Coded acoustic wave sensors and system using time diversity

    Science.gov (United States)

    Solie, Leland P. (Inventor); Hines, Jacqueline H. (Inventor)

    2012-01-01

    An apparatus and method for distinguishing between sensors that are to be wirelessly detected is provided. An interrogator device uses different, distinct time delays in the sensing signals when interrogating the sensors. The sensors are provided with different distinct pedestal delays. Sensors that have the same pedestal delay as the delay selected by the interrogator are detected by the interrogator whereas other sensors with different pedestal delays are not sensed. Multiple sensors with a given pedestal delay are provided with different codes so as to be distinguished from one another by the interrogator. The interrogator uses a signal that is transmitted to the sensor and returned by the sensor for combination and integration with the reference signal that has been processed by a function. The sensor may be a surface acoustic wave device having a differential impulse response with a power spectral density consisting of lobes. The power spectral density of the differential response is used to determine the value of the sensed parameter or parameters.

  12. Computationally Efficient Blind Code Synchronization for Asynchronous DS-CDMA Systems with Adaptive Antenna Arrays

    Directory of Open Access Journals (Sweden)

    Chia-Chang Hu

    2005-04-01

    Full Text Available A novel space-time adaptive near-far robust code-synchronization array detector for asynchronous DS-CDMA systems is developed in this paper. There are the same basic requirements that are needed by the conventional matched filter of an asynchronous DS-CDMA system. For the real-time applicability, a computationally efficient architecture of the proposed detector is developed that is based on the concept of the multistage Wiener filter (MWF of Goldstein and Reed. This multistage technique results in a self-synchronizing detection criterion that requires no inversion or eigendecomposition of a covariance matrix. As a consequence, this detector achieves a complexity that is only a linear function of the size of antenna array (J, the rank of the MWF (M, the system processing gain (N, and the number of samples in a chip interval (S, that is, 𝒪(JMNS. The complexity of the equivalent detector based on the minimum mean-squared error (MMSE or the subspace-based eigenstructure analysis is a function of 𝒪((JNS3. Moreover, this multistage scheme provides a rapid adaptive convergence under limited observation-data support. Simulations are conducted to evaluate the performance and convergence behavior of the proposed detector with the size of the J-element antenna array, the amount of the L-sample support, and the rank of the M-stage MWF. The performance advantage of the proposed detector over other DS-CDMA detectors is investigated as well.

  13. Gene-Auto: Automatic Software Code Generation for Real-Time Embedded Systems

    Science.gov (United States)

    Rugina, A.-E.; Thomas, D.; Olive, X.; Veran, G.

    2008-08-01

    This paper gives an overview of the Gene-Auto ITEA European project, which aims at building a qualified C code generator from mathematical models under Matlab-Simulink and Scilab-Scicos. The project is driven by major European industry partners, active in the real-time embedded systems domains. The Gene- Auto code generator will significantly improve the current development processes in such domains by shortening the time to market and by guaranteeing the quality of the generated code through the use of formal methods. The first version of the Gene-Auto code generator has already been released and has gone thought a validation phase on real-life case studies defined by each project partner. The validation results are taken into account in the implementation of the second version of the code generator. The partners aim at introducing the Gene-Auto results into industrial development by 2010.

  14. Selection of a computer code for Hanford low-level waste engineered-system performance assessment

    International Nuclear Information System (INIS)

    McGrail, B.P.; Mahoney, L.A.

    1995-10-01

    Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites

  15. New serial time codes for seismic short period and long period data acquisition systems

    International Nuclear Information System (INIS)

    Kolvankar, V.G.; Rao, D.S.

    1988-01-01

    This paper discusses a new time code for time indexing multichannel short period (1 to 25 hz) seismic event data recorded on a single track of magnetic tape in digital format and discusses its usefulness in contrast to Vela time code used in continuous analog multichannel data recording system on multitrack instrumentation tape deck. This paper also discusses another time code, used for time indexing of seismic long period (DC to 2.5 seconds) multichannel data recorded on a single track of magnetic tape in digital format. The time code decoding and display system developed to provide quick access to any desired portion of the tape in both data recording and repro duce system is also discussed. (author). 7 figs

  16. Coding and Billing in Surgical Education: A Systems-Based Practice Education Program.

    Science.gov (United States)

    Ghaderi, Kimeya F; Schmidt, Scott T; Drolet, Brian C

    Despite increased emphasis on systems-based practice through the Accreditation Council for Graduate Medical Education core competencies, few studies have examined what surgical residents know about coding and billing. We sought to create and measure the effectiveness of a multifaceted approach to improving resident knowledge and performance of documenting and coding outpatient encounters. We identified knowledge gaps and barriers to documentation and coding in the outpatient setting. We implemented a series of educational and workflow interventions with a group of 12 residents in a surgical clinic at a tertiary care center. To measure the effect of this program, we compared billing codes for 1 year before intervention (FY2012) to prospectively collected data from the postintervention period (FY2013). All related documentation and coding were verified by study-blinded auditors. Interventions took place at the outpatient surgical clinic at Rhode Island Hospital, a tertiary-care center. A cohort of 12 plastic surgery residents ranging from postgraduate year 2 through postgraduate year 6 participated in the interventional sequence. A total of 1285 patient encounters in the preintervention group were compared with 1170 encounters in the postintervention group. Using evaluation and management codes (E&M) as a measure of documentation and coding, we demonstrated a significant and durable increase in billing with supporting clinical documentation after the intervention. For established patient visits, the monthly average E&M code level increased from 2.14 to 3.05 (p coding and billing of outpatient clinic encounters. Using externally audited coding data, we demonstrate significantly increased rates of higher complexity E&M coding in a stable patient population based on improved documentation and billing awareness by the residents. Copyright © 2017 Association of Program Directors in Surgery. Published by Elsevier Inc. All rights reserved.

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

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. Health consequences and health systems response to the Pacific U.S. Nuclear Weapons Testing Program.

    Science.gov (United States)

    Palafox, Neal A; Riklon, Sheldon; Alik, Wilfred; Hixon, Allen L

    2007-03-01

    Between 1946 and 1958, the United States detonated 67 thermonuclear devices in the Pacific as part of their U.S. Nuclear Weapons Testing Program (USNWTP). The aggregate explosive power was equal to 7,200 Hiroshima atomic bombs. Recent documents released by the U.S. government suggest that the deleterious effects of the nuclear testing were greater and extended farther than previously known. The Republic of the Marshall Islands (RMI) government and affected communities have sought refress through diplomatic routes with the U.S. government, however, existing medical programs and financial reparations have not adequately addressed many of the health consequences of the USNWTP. Since radiation-induced cancers may have a long latency, a healthcare infrastructure is needed to address both cancer and related health issues. This article reviews the health consequences of the Pacific USNWTP and the current health systems ability to respond.

  20. An Embedded System for Safe, Secure and Reliable Execution of High Consequence Software

    Energy Technology Data Exchange (ETDEWEB)

    MCCOY,JAMES A.

    2000-08-29

    As more complex and functionally diverse requirements are placed on high consequence embedded applications, ensuring safe and secure operation requires an execution environment that is ultra reliable from a system viewpoint. In many cases the safety and security of the system depends upon the reliable cooperation between the hardware and the software to meet real-time system throughput requirements. The selection of a microprocessor and its associated development environment for an embedded application has the most far-reaching effects on the development and production of the system than any other element in the design. The effects of this choice ripple through the remainder of the hardware design and profoundly affect the entire software development process. While state-of-the-art software engineering principles indicate that an object oriented (OO) methodology provides a superior development environment, traditional programming languages available for microprocessors targeted for deeply embedded applications do not directly support OO techniques. Furthermore, the microprocessors themselves do not typically support nor do they enforce an OO environment. This paper describes a system level approach for the design of a microprocessor intended for use in deeply embedded high consequence applications that both supports and enforces an OO execution environment.